bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on Factors Affecting CHF Based on Factorial Analysis in Narrow Rectangular Channel Under Natural Circulation
Authors: Zichao Li , Tao Zhou , Shun Shi , Haider Amir , Bing Li , Zejun Xiao
Published: October 2018

Research on influencing factors of CHF in narrow rectangular channel under natural circulation is of great significance to the safety of reactors. Taking the narrow rectangular experimental device as the research object, influencing factors of CHF in narrow rectangular channel were experimentally studied under natural circulation. With factorial analysis, effects of different factors and their interactions on CHF were analyzed. It is found that the contribution rate of mass flow rate is the largest, followed by the effect of outlet dryness, followed by the effect of system pressure. Their interactions between different factors have little effects on CHF in narrow rectangular channel under natural circulation.
Transient Modeling of Advanced High Temperature Reactor (AHTR) in RELAP5/SCDAPSIM/MOD 4.0
Authors: Lin Hsun-Chia , Zhang Sheng , Shi Shanbin , Sun Xiaodong , Christensen Richard
Published: October 2018

The Advanced High Temperature Reactor (AHTR) is a fluoride-salt-cooled high-temperature reactor (FHR) design concept that is currently being developed at Oak Ridge National Laboratory for efficient production of electricity with improved safety features. Transient analyses of different scenarios are critical to demonstrate the safety of the AHTR design. An AHTR reactor model has been developed using RELAP5/SCDAPSIM/MOD 4.0. Thermodynamic and transport properties of three molten fluoride salts, namely FLiBe, FLiNaK, and KF-ZrF4, have been implemented into the RELAP5 code. The AHTR RELAP5 model consists of a reactor core, an upper plenum, a lower plenum, three primary loops, and three Direct Reactor Auxiliary Cooling Systems (DRACS) loops. DRACS Heat Exchangers (DHX) and Natural Draft Heat Exchangers (NDHX) are important components of DRACS and provide coupling between the primary loops and DRACS loops, and DRACS loops and air chimneys, respectively. Single-wall fluted tube heat exchanger designs have been proposed for the DHX and the NDHX to improve heat transfer performance in the two heat exchangers, and heat transfer correlations for fluted tubes have also been implemented into the RELAP5 code.

In this study, steady-state reactor normal operation and two transient scenarios are analyzed with the RELAP5 AHTR model. Based on a thermal hydraulics Phenomena Identification Ranking Table (PIRT) exercise, loss of forced circulation (LOFC) and loss of multiple DRACS loops are selected as the two transients for analysis. During transients, the decay heat is removed by the ambient air, fully relying on natural circulation/convection. The results of both transient scenarios show sufficient decay heat removal capabilities of DRACS with the proposed design.
Numerical and Experimental Investigation on Core Assembly Thermal-Gradient-Induced Deformation of Sodium-Cooled Fast Reactor
Authors: Ma Zehua , Wu Yingwei , Su G. H. , Tian Wenxi , Qiu Suizheng
Published: October 2018

In sodium-cooled fast reactor (SFR), thermal gradient is the paramount factor of assembly transient bowing, that may cause great reactivity change, accelerate wrapper vibration wear, hindering the motion of control/shutdown rods, or worse yet, threatening the integrity of assemblies. However, because of the complexity of multi-assembly contact and interaction problem, it is difficult to assess the impact of core deformation on reactor performance safety.

The Core Assembly Deformation Test Facility (CADTF) is designed to perform a series of thermal bowing tests by Xi‘an Jiao Tong University (XJTU) to investigate the core deformation behaviors under thermal gradient. In this paper, a finite element model was established to simulate the mechanical response of single assembly under different flat-to-flat thermal gradient. The single assembly restrained bowing test performed in CADTF is chosen to validate the model. In the model, the measured temperature distribution as well as temperature-dependent elastoplastic and thermal expansion properties were taken into consideration. To ensure the model reliability, iterative computation is conducted by adjusting the friction coefficient of the load pads to match the calculated and measured contact force. According to the results, it can be seen that the three-dimensional displacement of assembly shows relatively good agreement with the experimental data. Therefore, it can be concluded that the model is capable of performing core deformation analysis for SFR.
Study on Control Strategy of Hydrogen Risk in Small Containment Under Severe Accident
Authors: Zhiqiang Zou , Ming Zhang , Huanhuan Peng , Liqiang Hou , Chunrui Deng , Xuedong Qiao
Published: October 2018

Hydrogen combustion or detonation happened in the containment within the process of the small reactor severe accident may threaten the integrity of the containment. In this paper, based on systemic design of the Small Modular Reactor (SMR) surrounded by the steel containment, an innovatory combustible gas control strategy which using the passive containment cooling system (PCCS) and passive autocatalytic recombiners (PARs) is made to control the hydrogen risk in the small steel containment. A severe accident hydrogen risk analysis model is built by the integrative severe accident analysis program MELCOR, the validity of the strategy is analyzed at a typical severe accident. With this understanding, a three-dimensional computed fluid dynamics hydrogen behavior analysis model of the small steel containment is established by GASFLOW code, and the gas distribution non-uniformity in the containment is analyzed. The result shows that the steam condensation process in the containment could be slowed down by controlling the action of PCCS, and the steam concentration in the containment could be in the range of high level, while the oxygen concentration could be in the range of low level. If the PARs were added, the PARs could consume the hydrogen and oxygen in the containment sustainedly. The containment atmosphere could be in an inerted condition during the accident process, even though the hydrogen concentration in the containment is high. The gas distribution non-uniformity analysis result shows that oxygen concentration was low in the extent of high hydrogen concentration and high steam concentration, the steam, oxygen and hydrogen distribution non-uniformity would not affect the inerted atmosphere of containment.
The Role of the NRC in License Renewals
Authors: Miranda Samuel
Published: October 2018

So far, the licenses of more than 85% of US operating plants have been renewed, authorizing them to continue operations for an additional 20 years past the end of their original, 40-year operating licenses.

10 CFR §54, which governs the NRC’s renewal of plant operating licenses, defines the Current Licensing Basis (CLB), and its role in license renewal applications. The CLB includes, inter alia, compliance with all the regulations in 10 CFR §50, and its appendices, including the design-basis information presented in final safety analysis reports (FSARs). Appendix A of 10 CFR §50 specifies the General Design Criteria (GDCs), and defines the conditions under which they must be satisfied. For example, Appendix A defines anticipated operational occurrences (AOOs) as incidents that may occur during the lifetime of a particular plant. Two standards of the American Nuclear Society (ANS) [1] [2] redefined AOOs as incidents that may occur during the calendar year for a particular plant, and added a class of events that may occur during the lifetime of a particular plant.

The ANS standards defined a categorization scheme that puts all the various types of incidents into four categories (i.e., Conditions: I, II, III, and IV), and specifies the GDCs and other criteria that must be met for each of the categories. AOOs, or Condition II events, for example, must not result in any fuel damage. Licensees have committed to abide by the ANS categorization system, and to comply with all of the categories’ acceptance criteria. These commitments are in their Final Safety Analysis Reports (FSARs), which are part of their CLBs.

Conditions I, and II define relatively frequently occurring incidents, and require that their consequences to be benign. Condition III events, however, are limited to only a very few incidents during the lifetime of a plant. Their consequences are not benign. (Condition IV is the most severe category. Condition IV events are not expect to occur at all.)

Since the frequency of occurrence of Condition III events is defined in terms of plant lifetime, it follows that lengthening the plant lifetime (e.g., from 40 to 60 years) will lead to the occurrence of more Condition III events. These events can result in fuel damage, or worse. The CLB of a renewed license plant, therefore, will have to account for more Condition III events.

This paper focuses upon how Condition III events can affect the CLB during an extended plant lifetime. It also discusses the concept of extending plant operating licenses by 20, or 40 years, and its potential impact upon the public health and safety.
Uncertainty Analysis of Scaling Calculations From LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization Applying to a Four-Loop PWR
Authors: Kinoshita Ikuo
Published: October 2018

The Best Estimate Plus Uncertainty (BEPU) method is applied to analysis of the “intentional depressurization of steam generator secondary side” which is an accident management procedure in a small break loss-of-coolant accident with high pressure injection system failure. In the present study, scaling calculations from the LSTF small break LOCA tests were carried out for a conventional Westinghouse type four-loop PWR. The two test cases were selected with different break size and different depressurization conditions to ensure the reliability for the analyses for the accident scenario. The uncertainty propagation analyses were performed for the PWR. The dominating input uncertainty parameters, i.e. those few with the highest influence on the output uncertainty were identified. The analysis results were compared with those for the LSTF to address the scaling up capability and the similarity evaluation between the LSTF model and the PWR plant model. It was found that the PWR plant model results had overall agreements with the LSTF model, and the uncertainty of the predicted PCT included the measured PCT. Furthermore, the correlation coefficients between the input uncertainty parameters and the PCT for the PWR plant model had similarities with those for the LSTF model.
Research on Gamma Camera Imaging Characteristics
Authors: Zhang Quanhu , Zuo Wenming , Li Sufen , Hou Suxia , Zhuang Lin , Zhou Wenheng
Published: October 2018

Gamma camera imaging technology is a non-destructive passive radiation imaging technology, which can quickly find the unknown source location, search the exact number of radioactive sources and relative intensity. Therefore, it is very important and widely used in the fields of effective regulation of radioactive sources, handling of various nuclear emergencies, nuclear arms control and other fields. In the practical application of gamma camera, it often faces the imaging difference caused by the difference of radiation source intensity, detection time and detection distance. It is helpful to study the change of imaging characteristics under different experimental conditions for the practical application of gamma camera under different scenes. In this paper, the structure and imaging principle of gamma camera are analyzed in detail. Using the Cartogam portable gamma camera, a set of comparative experiments are carried out to study the time characteristics, distance characteristics and source intensity characteristics of the gamma camera. The results show that the imaging quality of gamma camera is positively correlated with the time source intensity, negatively correlated with the distance. For a milliCurie source, the gamma camera has very good fast-position resolution at a distance of 1 meter from the radioactive source and can form a more complete hot spot image within 5 minutes. When the distance becomes larger, the radioactive source needs at least 20 minutes to form a more accurate hot spot image. The hot spot is no longer as complete as a concentric circle structure, but can achieve precise positioning. For a strong source of more than ten milliCurie, immediate imaging within two minutes can be basically achieved within two meters. Under multi-source conditions, when the source intensities differ greatly and the distance between sources is relatively close, the detection of weak source can not be achieved by increasing the measurement time. However, by observing the counting images in a short period of time, the possibility of existence of a weak source can be deduced. Therefore, in the practical application of the gamma camera, it is necessary to constantly adjust its imaging conditions to ensure the detection of weak source verification. In this paper, the Monte Carlo model of gamma camera is set up to simulate the imaging. Compared with the actual imaging hot spots, the simulated images can correctly reflect the hot spot graph’s level distribution, which has the value of further research.
Modification of Filtered Air Intake Flowrate to Improve Control Room Habitability
Authors: Liu Xinjian , Shu Weipeng , Wang Mengxi
Published: October 2018

Control room habitability (CRH) shall be maintained to provide adequate protection for control room operators, such that they can remain in the control room envelope (CRE) safely for an extended period and thus control the nuclear facility during normal and accident conditions. A critical objective of CRH systems is to limit operator doses and/or exposure to toxic gases. The CRH systems does this by the combination of the intake of filtered air, isolation of outside air, recirculation systems and etc.

Among the parameters determining radioactivity in a control room (in proportion to radiation doses of operators), intake flowrate of filtered air is an important one. For different types of accident source terms, the evolution of operator doses in a control room versus intake flowrate were analyzed in this paper. It turns out that the increase of intake flowrate results in larger operator doses when inert radioactive gases are the dominant radioactive substances. On the contrary, increasing intake flowrate does good to lower the irradiation level of control room operators when radioactive aerosols dominate the source terms. The rationality behind this fact was interpreted in detail in this paper, with special attention paid to the unfiltered in-leakage rate. It can be inferred that an optimal intake flowrate probably exists leading to the minimum operator dose under an actual accident condition.

This paper then performed a calculation analysis based on design parameters and source terms of design basis accident of LOCA (a large break loss of coolant accident) accident. The evolution of operator dose was found to be a U-curve versus increasing intake flowrate, which proved the existence of the abovementioned optimal intake flowrate of filtered air for CRH systems. Furthermore, the sensitivity analysis of intake flowrate was carried out to study the effects of unfiltered in-leakage rate and filtered recirculation.

This study indicates that intake flowrate of filtered air can significantly influence the CRH. For different accidents, the intake flowrate should be properly modified rather than set as a fixed value. To optimize the radiological habitability of control rooms, the effects of unfiltered in-leakage must be taken into consideration. Besides, filtered recirculation is an effective way to control radiation exposure caused by iodine and radioactive aerosols.
Relationship Between Void Fraction and Electrical Characteristics in Gas-Liquid Two Phase Flow
Authors: Takakura Yuya , Jeon MinHo , Do MinJae , Kawashima Daisuke , Takamiya Sohei , Takei Masahiro
Published: October 2018

The purpose of this study is the calibration in normalization technique for the real-time measurement of void-fraction inside fluid by using VC system at each temperature condition. There is an existing method of measuring electrical response called a VC system. A VC system is a voltage applied current measured system. The VC method measures electrically between electrodes attached around the measurement subject and detects internal distribution from impedance difference of the component to be measured. We deal with water and air in this research. In this method, we measure the difference between the electrical characteristics of air and water. And the measurement technique is requirement of pre-measurement. In homogeneous case, we need to measure current for normalization with real-time measured current. This method has problem that the impedance of water depends on the temperature. So the impedance of water at pre-measurement is different as compared real-time impedance. That’s why, the technique for measuring void-fraction is not high accuracy at present. The measurement technology should be improved to obtain more accurate area of air inside fluid. This requirement is achieved by real-time measurement and real-time separation based on the different electrical properties of air and water while calibrating with temperature. In order to improve accuracy of the proposed method, we conducted a system comparison experiment before calibration with temperature and after calibration. In conclusion, we could measure the void fraction in real time more accurately.
Optimization of Moving Particle Semi-Implicit (MPS) Method and Studying of Flow Instability
Authors: Guo Kailun , Chen Ronghua , Qiu Suizheng , Tian Wenxi , Su Guanghui , Wu Junmei
Published: October 2018

Multiphase flow widely exists in the nature and engineering. The two-phase flow is the highlight of the studies about the flow in the vessel and steam explosion in nuclear severe accidents. The Moving Particle Semi-implicit (MPS) method is a fully-Lagrangian particle method without grid mesh which focuses on tracking the single particle and concerns with its movement. It has advantages in tracking complex multiphase flows compared with gird methods, and thus shows great potential in predicting multiphase flows. The objective of this thesis is to develop a general multiphase particle method based on the original MPS method and thus this work is of great significance for improving the numerical method for simulating the instability in reactor severe accident and two-phase flows in vessel. This research is intended to provide a study of the instability based on the MPS method. Latest achievements of mesh-free particle methods in instability are researched and a new multiphase MPS method, which is based on the original one, for simulating instability has been developed and validated. Based on referring to other researchers’ papers, the Pressure Poisson Equation (PPE), the viscosity term, the free surface particle determination part and the surface tension model are optimized or added. The numerical simulation on stratification behavior of two immiscible flows is carried out and results are analyzed after data processing. It is proved that the improved MPS method is more accurate than the original method in analysis of multiphase flows. In this paper, the main purposes are simulating and discussing Rayleigh-Taylor (R-T) instability and Kelvin-Helmholtz (K-H) instability. R-T and K-H instability play an important role in the mixing process of many layered flows. R-T instability occurs when a lower density fluid is supported by another density higher fluid or higher density fluid is accelerated by lower density fluid, and the resulting small perturbation increases and eventually forms turbulence. K-H instability is a small disturbance for two different densities, such as waves, at the interface of the two-phase fluid after giving a fixed acceleration in the fluid. Turbulence generated by R-T instability and K-H instability has an important effect in applications such as astrophysics, geophysics, and nuclear science.
Preliminary Hazard Analysis of Uranium Hexafluoride Accident
Authors: Lei Chen , JianGang Zhang , GuoQiang Li , ShuTang Sun , DongYuan Meng , Ning Wang
Published: October 2018

Uranium hexafluoride (UF6) accident is a typical accident in the nuclear fuel cycle. It combine radioactive and chemical hazards, so it is necessary to attach great importance to the UF6 accident. This paper analyzed and summarized the possible accident scenarios, causes and consequences, and classified the UF6 accident risk factors, and put forward corresponding preventive and emergency measures. Preliminary hazard analysis of the accident can help us better understand the accident process, so we can takes steps for corresponding risk factors in advance, and prevent it will not be developed into an accident, so than we can obtain the effect of nip in the bud.
Preliminary Neutronics and Thermal-Hydraulics Study on Thorium-Based HTR-PM With Outer Breeding Zone
Authors: Wang Qiudong , Xia Bing , Guo Jiong , She Ding , Shi Lei , Zhang Zuoyi
Published: October 2018

In this work, a two-zone reactor core, which contains an inner driving zone and an outer ThO2 breeding zone, is designed under the framework of the HTR-PM. The main aim of this work is to investigate the feasibility of thorium utilization in the mature design of the HTR-PM with the inherent safety features. The neutronics and thermal-hydraulics characteristics are investigated to optimize the design parameters by using VSOP. The aim of optimization is to maximize the conversion of thorium to 233U in the breeding zone. The preliminary results indicate that the volume ratio of the breeding zone to the driving zone has significant influence on the power peaking factor and the maximum fuel temperature in normal operation and accidental conditions. On the other hand, the increase of reactor power will lead to increase of maximum fuel temperature after DLOFC accident. More heavy metal loading in the breeding zone will raise 233U yield, while the influence of fuel particle radius on the conversion ratio is negligible. An optimized 200 MWt two-zone reactor design is obtained with volume ratio of the driving zone to the breeding zone of 4:1, and 7 g and 30 g heavy metal per fuel sphere in the driving zone and the breeding zone, respectively.
Enhancement of Stochastic Sampling Capability in RMC Code
Authors: Shi Guanlin , Qiu Yishu , Wang Kan
Published: October 2018

As people pay more attention to nuclear safety analysis, sensitivity and uncertainty analysis has become a research hotspot. In our previous research, we had developed an integrated, built-in stochastic sampling module in the Reactor Monte Carlo code RMC [1]. Using this module, we can perform nuclear data uncertainty analysis. But at that time the uncertainty of fission spectrum was not considered. So, in this work, the capability of computing the uncertainty of keff induced by the uncertainty of fission spectrum, including tabular data form and formula form, is implemented in RMC code based on the stochastic sampling method. The algorithms and capability of computing keff uncertainty induced by uncertainty of fission spectrum in RMC are verified by comparison with the results calculated by the first order uncertainty quantification method [2].
Technical Insights of SSR-2/1 Safety of Nuclear Power Plants: Design (Rev.1)
Authors: Hua Zheng , Shuhong Wei
Published: October 2018

In 2016, IAEA revised and issued SSR-2/1 Safety of Nuclear Power Plants: Design and issued SSR-2/1 (Rev.1). Major revision in SSR-2/1 (Rev.1) is analyzed (especially new requirements after Fukushima nuclear accident). Problems in SSR-2/1 (Rev.1) are also discussed, which could be used as a reference in HAF102 revision and SSR-2/1 (Rev.1) application.
Numerical Research on Melt Pool Flow Characteristics Under Rolling Condition
Authors: Luo Simin , Wang Xin'an , Zhang Yapei , Zhang Dalin , Qiu Suizheng , Su Guanghui
Published: October 2018

In-vessel-retention (IVR) has become an important subject of severe accident mitigation strategy. Up to now, many experimental and numerical investigations have been performed on the natural convection characteristics in melt pools with volumetric heating. But these studies are limited to the melt pools under static condition. As floating nuclear reactors become increasingly popular among both commercial and military ships, for successful application of IVR in this occasion, research should be done on the heat transfer characteristics of melt pool under moving conditions. Currently, the specially-designed facility is under construction in Xi'an China for the relevant experiment and numerical studies are performed beforehand. In this paper, a hemisphere with an inner radius of 0.5m, similar to LIVE experimental facility, is chosen to simulate the melt pool. Its flow behaviors under periodic rolling condition are simulated by means of CFD calculation. This work may cast a light on the melt pool characteristics under moving conditions and could be further evaluated by future experimental data.
Impairment of Local Heat Transfer of the Turbulent Mixed Convection in a Vertical Flat Plate
Authors: Chae Myeong-Seon , Chung Bum-Jin
Published: October 2018

The heat transfer of the buoyancy-aided turbulent mixed convective flow in a vertical flat plate was investigated experimentally. Mass transfer experiments were carried out based on the heat and mass transfer analogy. The Rayleigh numbers ranged from 1.69 × 108 to 2.11 × 1013, depending on the height of the vertical flat plate. The Reynolds numbers varied from 4,585 to 17,320 for turbulent regimes. The test results for turbulent forced convections agreed well with the forced convection correlations established by Petukhov et al. The local heat transfer rates of the turbulent mixed flow exhibited the impairment of heat transfer compared to the forced convection and non-monotonous behavior along the axial position due to buoyancy effect. The local minimum heat transfer was 38.6% lower than the forced convection heat transfer. The turbulent mixed convection heat transfer is affected by the height of vertical plate.
An Instrument Established for the High Temperature Measurement of Ultraviolet-Visible Absorption Spectra of Molten Fluoride Salt Behaving As Coolant in the Molten Salt Reactor
Authors: Liu Hongtao , Liu Yiyang , Su Tao
Published: October 2018

Molten salts were widely used in nuclear and solar power field due to the excellent heat transfer and storage. Molten fluoride salts were selected as primary and secondary coolants in the Molten Salt Reactor Experiment (MSRE) developed by Oak Ridge National Laboratory (ORNL). Therefore, it is dramatically important to study the physical and chemical properties of molten fluoride salts that impact on the design of reactor core and thermohydraulics. The molecular structure directly determines the physical and chemical properties of matter, so it is also essential to study the structure of molten salts. Spectroscopy has been proven to be a very useful tool for investigating molten salts structures. However, the standard instrument is inapplicable for measurement of the high temperature molten salts, especially for molten fluoride salts. To obtain the ultraviolet-visible (UV-Vis) absorption spectra of molten salts at high temperature, an instrument was designed to study the structures of molten salts in situ. The instrument is mainly composed of a vertical pit furnace connecting with a glovebox and an assembled cuvette which can operate from room temperature up to 800°C. The assembled cuvette is made of Hastelloy C/N as the main body with a reverse ‘T’ contour and diamond or crystalline CaF2 etc. as the window plates, so it can withstand the corrosion produced by the sample and allow the interest light passing through. The effective spectral range of this instrument is from 200 to 1000 nm. Performances of the instrument are testified by spectral studies on water under room temperature and molten salts under high temperature.
A New Method of Integrating the RELAP5 to the RINSIM Simulation Platform
Authors: Tan Chao , Quiroga Victor M. , Fu Zheng , Xie Zhengquan
Published: October 2018

This paper introduces a method to control the calculation progress of the RELAP5 codes and integrate them with other codes by interacting boundary data (such as general tables, TDVs and TDJs) at each step.

This work is basically finished with the support of the RINSIM simulation platform. The paper gives a brief introduction on RINSIM that how it controls the codes progress, sends the codes control commands, shares the values of different codes’ common blocks or modules.

However, the work can’t be done by just using the RINSIM, it also needs to modify RELAP5 codes. With the codes’ modification of commands responding, data reading/writing interface, data interacting interface, time step control and so on, we can build interface subroutines to integrate codes onto the RINSIM.

At the end, the paper gives out the result of a transient calculation with an advanced PWR model. Compared to some old integration method, the new method has far more strong stability. And the result shows that the integration progress of the code does not obviously affect the calculation accuracy, but definitely extends the application fields because of the multiple functions supplied by the RINSIM.
Measurement of Dry-Out Heat Flux (DHF) of Debris Beds Using a Non-Heating Experimental Methodology
Authors: Moon Je-Young , Chung Bum-Jin
Published: October 2018

The dry-out heat flux (DHF) for the copper sphere beds was measured using the hydrogen generated at a high potential in the copper electroplating system of mass transfer. The uniform self-heating condition was simulated by the experimental methodology. The experimental apparatus consisted of a power supply, a data acquisition system, a polycarbonate cylinder containing the copper cathode beds and copper anode submerged in an aqueous solution of H2SO4. This study reported the experimental results for the bottom flooding condition with the 6 mm copper beds. The DHF data was obtained for the bed height of 20, 40 and 60 mm. The tendency of DHF on the bed height is similar to that of existing heat transfer studies. However, the absolute DHF value was quite different. We inspected the potential of the non-heating experimental method and reviewed further studies for reducing the discrepancy.
Evaluation of Explosivity From Fuel Coolant Interaction Experiments in the TROI Facility
Authors: Hong Seong-Wan , Park Rae-Joon
Published: October 2018

The explosivity of corium is evaluated from the fuel coolant interaction experimental data produced in the TROI facility.

From 62 experimental datasets used to simulate the fuel coolant interaction under a partial flooded condition of the reactor, the explosivity when observing the dynamic pressure and/or force measured in experiments is qualitatively evaluated depending on the coolant depth, the shape of the interaction chamber, the free fall height of the melt jet, and the water temperature effect. For 12 of the experimental datasets produced under the OECD/SERENA Project, the explositivity is quantitavely evaluated using the conversion ratio. The conversion ratio is also evaluated for three experimental datasets to simulate the fuel coolant interaction under reactor flooded conditions.

The corium system showed a relatively low explosivity compared to the ZrO2 or Al2O3 system. It also turned out that the flooding condition of the reactor cavity does not affect the change in explosivity.
Epistemic Uncertainty Quantification of Floor Responses for a Nuclear Reactor Building
Authors: Choi Byunghyun , Nishida Akemi , Li Yinsheng , Muramatsu Ken , Takada Tsuyoshi
Published: October 2018

After the 2011 Fukushima accident, engineers of nuclear power plants are looking beyond the basic design requirements and ensuring that countermeasures are built in to avert possible nuclear accidents. In seismic probabilistic risk assessment (SPRA), uncertainties can be classified in two ways as aleatory uncertainties or epistemic uncertainties. To improve the reliability of SPRA, the difference in seismic response due to difference of building modelings related to epistemic uncertainty was focused on. Two modeling methods were used for a seismic response analysis: a three-dimensional finite-element model and a conventional sway-rocking stick model. Simulated input ground motions related to aleatory uncertainty were generated for the input waves. Then, the seismic floor response results of the various input ground motions of the two modeling methods were quantified. For the uncertainty quantification related to the different building modelings, a statistical analysis of the floor response results of the nuclear reactor building were further performed. Finally, for the quantification of the uncertainty in the fragility analysis for SPRA, the way to use of these results were discussed.
Experimental Research on Heat Transfer Characteristic of Liquid Lead-Bismuth Eutectic Flowing in Annular Channel
Authors: Zhu Fengjie , Wu Junmei , Shi Leitai , Su G. H.
Published: October 2018

Liquid lead-bismuth eutectic (LBE) has been studied as a new type of coolant for an accelerator driven sub-critical system (ADS). And the gas-lift pump has been used to enhance the circulation capacity of coolant in ADS instead of mechanical pumps due to its simpler mechanical structure and higher security. The purpose of this experiment is to study the circulation capacity enhancement by gas-lift technique and the flow and heat transfer characteristic of liquid LBE in an annular channel. The experimental results show that: gas-injection can significantly increase liquid LBE mass flow rate, but the growth of liquid LBE mass flow rate will be reduced when gas flow rate reaches a value; The friction coefficient of liquid LBE in an annular channel decreases with the increase of Re and is larger than that calculated by Blasius formula at the same Re. For the convection heat transfer of liquid LBE in an annular channel, the heat conduction term is dominant, and Nusselt number increases with the increase of Peclet number. The experimental correlations of friction coefficient and convection heat transfer of liquid LBE in annular channel were fitted based on experimental data and compared with those from literature.
A Demonstration of Practical Elimination of Early or Large Fission Product Release for the UK ABWR Generic Design Assessment
Authors: Ang Ming Leang , Mohamud Nuh , Chitose Hiromasa , Hirokawa Naoki , Kimura Ryusuke
Published: October 2018

The demand for continuous improvement in safety of nuclear power plants has led to an international expectation that early or large releases of fission products as a result of severe accidents be practically eliminated for new reactor designs. The UK Department for Business, Energy and Industrial Strategy has recently published the Seventh UK Report which provided confirmation of UK demonstrating compliance with the obligations of the Convention on Nuclear Safety [Ref-1]. Relating to Article 14 on the Assessment and Verification of Safety, the UK nuclear Regulator (Office for Nuclear Regulation (ONR)) has the expectation of the demonstration of ‘practical elimination’ of potential severe accident states be included in the safety cases for new nuclear power plants [Ref-2]. In order to achieve this, the safety case should show either that it is physically impossible for the accident states to occur or the states can be considered to be extremely unlikely with a high degree of confidence by design provisions [Ref-3] [Ref-4] [Ref-5]. A demonstration framework was developed and applied successfully in the UK ABWR Generic Design Assessment (GDA) Pre-Construction Safety Report (PCSR) which was submitted to the ONR in August 2017 [Ref-6].

A summary of this demonstration is provided in this paper.
Study on Flow Structure in a Supersonic Steam Injector
Authors: Kamata Yuki , Fujishiro Masaya , Kaneko Akiko , Abe Yutaka
Published: October 2018

Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant.

Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model.

The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.
Evaluation of Jet Impact Region and Fluid Force Generated From Ruptured Pipes: Part 4 — Numerical Evaluation of Affected Region by Flashing Jet Flow
Authors: Morita Ryo , Watanabe Shun , Takahashi Shiro , Takamura Noriyuki
Published: October 2018

Nuclear power plants are designed to prevent the damage of safety installations and human safety due to the jet impingement when a pipe is ruptured. We have investigated evaluation methods for the design basis of protection of plants against effects of postulated pipe rupture using computational fluid dynamics analysis (CFD). The flashing steam jet was evaluated in this study. Shapes of flashing jet were obtained by CFD calculations, and we have found that the affected region was wider than usual steam jet flow and strongly depends on the nozzle exit flow conditions. This is thought as the different tendency from established standards.
Conceptual Core Design of HAPPY200 Reactor
Authors: Li Xiaosheng , Li Linsen , Peng Lianghui , Chen Xiaosong , Meng Zhaocan , Chen Yaodong
Published: October 2018

The pressure and coolant temperature of Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY200) is significantly lower than PWR of the NPP, the core design and analysis were completed according to the design parameters and features of HAPPY200. The fuel assembly and its feature was firstly designed and studied based on the investigation of different types of fuel assemblies. Then the core configuration was studied and optimized according to the design parameters of HAPPY200; Eventually, neutronics calculation of the core was performed and key parameters were obtained including cycle length, power distribution, control rod worth, reactivity coefficients and etc. The study shows that with the core design HAPPY200 can be operated for 18 months in full power and reactivity control system can maintain criticality of the core in the full cycle. Due to the non-soluble boron design of the reactivity control scheme, moderator temperature coefficient and isothermal temperature coefficient are both negative, the Doppler temperature coefficients and power coefficients in different phase of the lifetime and in different power levels are also negative, therefore, the reactivity safety of the reactor core can be ensured.
Numerical Research on Shock Resistance Safety Analysis of Ship Power Plant Valve
Authors: Wu Jun , Bai Fan , Liu Yong , Lao Xingsheng , Dai Chunhui
Published: October 2018

In order to increase the operation safety of ship nuclear power platform. A typical butterfly valve with nuclear safety class 2 of ship power plant is studied in this paper. Numerical analysis on shock resistance safety of butterfly valve is conducted by three-dimensional finite element method. Firstly, the load type under working condition and the stress evaluation criteria of valve safety analysis are determined based on national standards and ASME standards. Secondly, the finite element model of butterfly valve is constructed and the equivalent stress distribution of butterfly valve is obtained under static loads and constraint conditions. Moreover, the butterfly valve natural frequencies and vibration shapes are obtained by modal analysis under inspection working condition. Furthermore, the butterfly valve stress distributions on three coordinate directions are analyzed by shock response spectrum analysis according to the modal analysis results. Lastly, the total stress distribution of butterfly valve is obtained by superposition procedure. And the butterfly valve structure shock resistance safety is checked based on the stress evaluation criteria.
Fundamental Study on Development of Air-Coupled Ultrasonic Imaging Measurement for Fuel Debris Inspection
Authors: Kiuchi Hironobu , Kimura Shun , Hamdani Ari , Takahashi Hideharu , Kikura Hiroshige , Sasa Daisuke , Ohmori Shuichi
Published: October 2018

Retrieval of fuel debris is one of the most urgent issues regarding the decommissioning of Fukushima Daiichi Nuclear Power Plant (NPP). A retrieval of fuel debris requires a measurement system which can be applied in the severe environment to investigate inside the Reactor Pressure Vessel (RPV) and the Control Rod Drive (CRD) housing. In this study, an air-coupled ultrasonic measurement is proposed because of the feasibility of its application in the severe environment such as the high-radiation atmosphere. However, compared with using it in water or solids, there are some difficulties with applying the air-coupled ultrasonic technique. Therefore, we employed a point-focused transducer which can converge ultrasonic energy in a small area. As a fundamental study, this paper focuses on the experimental assessment of the potential feasibility and property of the developed air-coupled ultrasonic imaging. Firstly, to obtain the property and characteristics of the point-focused transducer, the sound field of the transducer was measured. The sound field showed the convergence rate, acoustic intensity and the beam width of the transducer. Secondly, in order to improve Signal to Noise ratio (SN ratio), the ultrasonic imaging method using a frequency correlation was designed taking the feeble ultrasonic echoes from complex shape object into consideration. Finally, the shape of a test specimen, which imitates a fuel debris, was measured with the developed imaging system. The ultrasonic image of the specimen showed a good agreement with a photo. In conclusion, the imaging using air-coupled point-focused ultrasound was developed to successfully visualize a mock fuel debris from a relatively short distance of 40 mm – 100 mm.
Tests of the Package for the Transport of Natural Uranium Hexafluoride
Authors: Hongchao Sun , Guoqiang Li , Dajie Zhuang , Shutang Sun , Dongyuan Meng , Yiren Lian , Lei Chen , Jiangang Zhang
Published: October 2018

Several tests has been conducted to illustrating the safety performance of a type of package for the transport of natural uranium hexafluoride meet the requirements of GB11806-2004 (Regulations for the safe transport of radioactive material). The requirements of GB11806-2004 are same with the requirements of IAEA SSR6 (Regulations for the safe transport of radioactive material). These tests include heat, cold, reduced external pressure, increased external pressure, free drop, thermal test. Certification testing was performed on full-scale model and a test plan was developed that identified the specific free drop necessary to evaluate both GB11806 and SSR-6 requirements. A total of two 0.6-m free drops were performed. The leaks were detected after each free drop test. The accelerations were recorded for use in finite element structural analyses. This paper reviews the test planning and results with a discussion of how the test and finite element structural analyses were combined.
Development of Single Pebble Benchmark Ex. I-2A for IAEA UAM CRP With MOOSE
Authors: Niu Jinlin , Wang Lidong , Guo Jiong , Li Fu
Published: October 2018

Ex. I-2a is Uncertainty Analysis in Modelling (UAM) of stand-alone thermal-hydraulics in steady state. It belongs to IAEA Coordinated Research Project (CRP). In this paper, the work is divided into two steps. Firstly, a single pebble model is developed based on Multiphysics Object Oriented Simulation Environment (MOOSE). The model includes two regions: homogeneous fuel region and shell region. One physical field is concerned: solid temperature field. The field is solved in a Jacobian-free Newton–Krylov (JFNK) way in the MOOSE framework. Secondly, perturbation of thermal parameters is done. In this paper, helium coolant temperature, convective heat transfer coefficient, power density and solid conductivity are the main thermal parameters which are concerned. Different kinds of perturbation of thermal parameters are conducted. Thus, variation of temperature profile across the pebble, Doppler temperature, moderator temperature and maximum temperature could be got.
Coupling Dependence of Multiple Operating Parameters on Burnup Credit Calculations for BWR Spent Fuel Assemblies
Authors: Wu Shang-Chien , Chao Der-Sheng , Liang Jenq-Horng
Published: October 2018

This study aims to investigate the coupling dependence resulting from three and four operating parameters for burnup credit calculations in boiling water reactor (BWR) spent fuel assemblies. Four operating parameters are under investigation, including fuel temperature, axial burnup profile, axial moderator density profile and control blade usage. In this study, the effects of variation on the curve of effective multiplication factor (keff) versus burnup (B) resulting from one and multiple operating parameters were defined as the single and compound effects, respectively. Particularly, the compound effects adopt more practical operating parameters than single effects does and thus affect the precise assessment to some extent. In our previous study, the compound effects resulting from two operating parameters were investigated in depth. However, the influence of compound effects resulting from three and four operating parameters on burnup credit calculation is still unknown. Therefore, this constitutes the purpose of this study. All the calculations were performed using SCALE 6.1 computer code together with the ENDF/B-VII 238 energy group neutron data library. Two geometrical models were established to represent the typical GE14 10 × 10 BWR fuel assembly and the GBC-68 storage cask. The results revealed that the reactivity deviation (or changes of keff, Δk) resulting from the compound effects was not a summation of the Δk’s resulting from the associated single effects. Moreover, such Δk discrepancies increase as B increases. In this study, the curves of keff versus B due to single and compound effects were approximated by a second degree polynomial of B. A general formula was thus proposed to express these curves.