bluuebunny/crossref_metadata_2025_split huggingface dataset:
CAP1400 Fuel Assembly Development and PWR Fuel Assembly Technology Progress in China
Authors: Zhu Libing , Ding Jie , Zhou Qin , Zhou Yunqing , Gan Fujun , Lu Junqiang , Zheng Yixiong , Li Jiwei , Ding Yang , Zeng Qifeng , Liu Song 
Published: October 2017

In 1990s the first generation of PWR fuel assembly FA300 was developed in China. With constantly design improvements over the following 20 years, three types of FA300 fuel assemblies have been developed with max fuel assembly average burnup reach 40000 MWd/tU and successfully used in 300MWe PWRs. By now over 1100 fuel assemblies of FA300 series have been successfully operated in three different 300MWe PWRs with satisfied operation reliability.
The CAP1400 fuel assembly development program was launched in 2010 which includes new zirconium cladding material development, UO2 pellet development, high performance fuel assembly mechanical structure development, fuel rod performance code and fuel assembly seismic analysis code development, out-of-pile fuel assembly test facility construction and in core irradiation program. The main purpose of CAP1400 fuel assembly development program is to meet the needs of self-sufficient fuel supply to CAP1400 reactor.
Based on FA300 and CAP1400 fuel assembly development, a full series of PWR Fuel R&D technology and test system has been successfully established by SNERDI which will continuously support the fuel assembly improving and new type of fuel development. This paper will mainly describe the PWR fuel technology including cladding technology, pellet technology, fuel assembly mechanical design technology, fuel assembly test facility and test technology, fuel rod and fuel assembly code development progress.
Furthermore, the development progress of CAP1400 fuel assembly will also be introduced. By now two types of new zirconium alloys have been selected as candidate alloys, all of the out-of pile performance tests have been finished. Both two alloys show good corrosion resistance. Through full series of fuel assembly components performance tests, the fuel assembly has finished design finalization. Most of the fuel assembly mechanical and hydraulic tests will be finished by the end of 2016. The test reactor irradiation program and commercial PWR irradiation program are also on the schedule. After the LTA program and commercial application licensing, the CAP1400 fuel assembly is anticipated to provide adequate burnup capability and operation reliability to CAP1400 reactor in China.
Thermal-Hydraulic Analysis of the EBR-II SHRT-45R Unprotected Loss of Flow Experiment With Modified RELAP5
Authors: Simiao Tang , Jian Song , Dalin Zhang , Chenglong Wang , Suizheng Qiu , Guanghui Su , Wenxi Tian 
Published: October 2017

Sodium-cooled fast reactor (SFR) is one of most promising Generation IV reactor technology and has a rapid development in recent years. Experimental Breeder Reactor II (EBR-II) designed by Argonne National Laboratory (ANL) is a typical sodium-cooled fast reactor with a sodium-bonded metallic fuel core, featured with reactor negative reactivity feedback. In order to verify and validate (V&V) the inherent safety performance of sodium-cooled fast reactor, the International Atomic Energy Agency (IAEA) established the coordinated research project (CRP) in which Xi’an Jiaotong University has participated. The “Benchmark Analysis of EBR-II shutdown heat removal tests” was conducted by ANL. SHRT-45R, namely Unprotected Loss of Flow (ULOF), is one of the reactor tests among many transients. Argonne National Laboratory has prepared a detailed benchmark specification and has provided the enough benchmark data for SHRT-45R. At the beginning of SHRT-45R both primary main pumps and the intermediate loop pump were synchronously tripped to simulate the unprotected loss of flow accident. During the test, the plant protection system (PPS) was disabled to initiate a control rod scram. The SHRT-45R test demonstrated that EBR-II could keep in safe during the potentially adverse consequences of unprotected accidents.
This paper introduces the models for predicting SHRT-45R in detail and presents the results of the analysis of the Unprotected Loss of Flow (ULOF) test SHRT-45R performed in the EBR-II reactor. The thermal-hydraulic calculations are performed with Modified RELAP5 in which the thermodynamic and transport properties of liquid and vapor state sodium have been supplemented, as well as the specific heat transfer correlations. The numerical results show that the loss of forced coolant flow causes the coolant temperatures in the instrumented subassemblies XX09 and XX10 to increase to a peak point but keep at an acceptable level about 930K and 850K at the early state of accidents, and then the reactor can shut down by itself due to the negative reactivity feedback including Doppler reactivity feedback and density reactivity feedback after 70s. The variation of the key thermal-hydraulic parameters including the coolant temperatures and fuel cladding temperature in the instrumented subassembly has a good agreement with the experimental data in general. The results could not only verify and validate the inhere safety performance of sodium-cooled fast reactors during potentially unprotected accidents but also demonstrate the analysis capability of the modified RELAP5 for sodium-cooled fast reactors.
Conceptual Core Design of Breeding BWR
Authors: Guo Rui , Yamaji Akifumi 
Published: October 2017

High breeding with light water cooling is not easy to be achieved. The main obstacle is the moderating effect of light water, which softens the neutron spectrum. Decreasing the volume ratio of coolant to fuel is normally introduced as a way to harden the neutron spectrum and achieve breeding with light water cooling. Therefore, the tight-lattice assembly was proposed to design reactors cooled by light water with hard neutron spectrum. However, most of them were High Conversion LWRs and none achieved high breeding to meet the growth rate of energy demand in advanced countries.
Tightly packed fuel assembly is designed for the purpose of high breeding. The number ratio of hydrogen atoms to heavy metal atoms (H/HM) in this assembly is significantly reduced to less than 0.1 which is about 1/6 of that of Reduced-Moderation Water Reactor (RMWR). Super Fast Breeding Reactor (Super FBR) is one kind of Supercritical Light Water Cooled Reactors (Super LWRs), which adopts these assemblies, obtaining high breeding of CSDT (less than 50 years).
The high breeding performance of Super FBR indicates that, application of the tightly packed fuel assembly on conventional LWR-type reactors, such as BWR-type or PWR-type reactor, may also be effective in achieving high breeding. Compared with Super FBR, the conventional LWR-type reactors with technologies which are currently in use are expected to be easier to implement. When comparing the two main LWR types, BWR-type and PWR-type, BWR-type gains more advantages on breeding, since the coolant is boiling water that generates larger amount of void in the reactor core, leading to a harder neutron spectrum. Meanwhile, from the viewpoint of safety, the negative void reactivity should be satisfied, which is consistent with conventional LWRs. From the viewpoint of neutron economy, high enrichment should be avoided as well.
This study aims to design the BWR-type reactor with the tightly packed fuel assemblies, which attains both high breeding and negative reactivity.
Development of 4S BEPU Method and Application in CAP1400 Containment Response Analysis
Authors: Xin Liu , Benxue Hu , Zhe Wang , Guodong Wang , Zhangli Wang , Di Zhang , Ping Yang 
Published: October 2017

Compared with conservation evaluation model, best estimate plus uncertainty (BEPU) method can obtain more realistic results and gain larger license margins with respect to the safety criteria. In view of this, a BEPU method named 4S (SNERDI Statistical Solution for Safety) has been developed, according to the basic principles of evaluation model development and assessment of RG 1.203.
The characteristics of 4S method are as follows: The output uncertainty is quantified by using random sampling and propagation of input uncertainties. Global sensitivity analysis is used to support PIRT establishment. Uncertainties of model parameters are calibrated and validated by using separate effects tests considering measuring uncertainties. DAKOTA code is used for uncertainty and sensitivity analysis. An automatic BEPU analysis platform has been developed by coupling DAKOTA and different reactor safety analysis codes, and code calculations can be performed in parallel.
BEPU analysis of mass and energy release and containment pressure response of CAP1400 under a postulated double-ended cold leg break loss of coolant accident (DECL LOCA) has been carried out by coupling DAKOTA, a mass and energy release analysis code and a containment analysis code. In total, 21 uncertain input parameters are considered. To make the results more stable, the sample size is 124 and the third highest peak pressure is used as the pressure upper bound (with 95%/95% probability/confidence) based on Wilks’ formula. The calculated results show that the peak pressure upper bound is obviously lower than the present conservation method used in license application, with more than 10% analysis margin. Influences of input parameter uncertainties on the containment peak pressure have been analyzed, according to the partial rank correlation coefficients calculated by DAKOTA. The results show that the input parameters mainly affecting the peak pressure are the containment condensation heat transfer multiplier, initial containment temperature, break resistance, decay heat, initial containment pressure, Core Makeup Tank (CMT) resistance multiplier and initial containment humidity.
Study of Flow and Heat Transfer of Low Power Natural Circulation Under Rolling Condition
Authors: Ji Peng , Gao Puzhen , Zhang Yinxing 
Published: October 2017

Experimental and numerical studies on the flow and heat transfer characteristics of low heat flux natural circulation under the rolling condition were conducted in this paper.
The flow loop applied in the experiment is fixed on a rolling platform and the test fluid is deionized water. The testing channel made of stainless steel is narrow rectangular with the cross section of 2×40 mm2 and the heat length is 1100 mm. Enough experimental data points are obtained for a system pressure range of 0.1–0.3 MPa and an inlet temperature range of 50°C∼80°C. The testing channel is heated by electric, and the heat flux is less than 60 kW/m2. The rolling platform is driven by the crank rocker mechanism, and the rolling angle of the experimental loop varies with sine function. Driven by low pressure head, the low heat flux natural circulation has a negative flow value in the trough because of the effect of additional pressure drop under rolling condition. The experimental results show that the temperature fluctuation of the coolant is more severe because of intermittent reverse of the flow, and the parameter of outlet water will turn high under the effect of repeated heating. At the same time the amplitude of the temperature fluctuation in the outlet and inlet increase sharply compared with the condition that the flow doesn’t reverse. Under this condition the temperature of water increases as the additional inertia force raises because the level of flow reverse is improved and the working fluid is heated for more than once.
The heat transfer of low power natural circulation under rolling condition is discussed in this paper. The analysis finds that the rolling condition doesn’t strengthen the heat transfer absolutely when the heat power is low. The average heat transfer coefficients of the natural circulation under rolling conditions increase faster than those under steady state when strengthening the power of heat. Considering the flow will reverse periodically, the reason why the heat transfer is weakened at low power condition is given.
Cyber Security Assessment of Component Off-the-Shelf Based NPP I&C System Using IMECA Technique
Authors: Zelinko Ilona , Kharchenko Vyacheslav , Leontiev Konstantin 
Published: October 2017

Nowadays cyber security assurance is one of the key challenges of safety critical software based NPP I&C (Nuclear Power Plants Instrumentation and Control) systems requirements profiling, development and operation. Any I&C system consists of a set of standard software (SW), hardware (HW) and FPGA components. These components can be selected and combined in different ways to address the particular control and safety assurance related tasks. Some of them are proprietary software (PS) and commercial off-the-shelf (COTS) components developed previously. Application of such components reduces the level of safety and cyber security, because they can contain vulnerabilities that were created intentionally. In this case, targeted attacks can lead to a system failure.
National Vulnerability Database (NVD) and other open databases contain information about vulnerabilities which can be attacked by insiders or other intruders and decrease cyber security of NPP I&C systems.
In this paper, we propose a safety assessment technique of NPP I&C systems, which consists of the following procedures:
1. Analysis of I&C architecture to assess influence of OTS component failures on dependability (reliability and safety) of the system. For that purpose, FMEDA or similar techniques can be applied.
As a result, three-dimension criticality matrixes (CM) (with metrics of detection, probability and severity) are developed for different components (SWFCM and HW/FPGAFCM).
2. The IMECA-based assessment of OTS components and their configuration. In this case, CMs (SWICM and HW/FPGAICM) describe the degree of failure component influence on cyber security.
3. Joining of criticality matrixes (SWFCM and HW/FPGAFCM, SWICM and HW/FPGAICM), impact analysis of components depending on degree of influence on cyber security and safety as a whole.
4. Developing of Security Assurance Case and selecting of countermeasures according to safety (cyber security)/costs criteria.
The developed tool supports creation of criticality matrixes for each analyzed component of the system and I&C as a whole. Joining of criticality matrixes allows creating common matrix for system cyber security and functional safety. The tool supports decision making to optimize choice of countermeasures according to criterion of safety and security/cost criterion.
RELAP5 STH and Fluent CFD Coupled Calculations of a PLOHS + LOF Transient in the HLM Experimental Facility CIRCE
Authors: Angelucci M. , Martelli D. , Forgione N. , Tarantino M. 
Published: October 2017

This work describes the activity performed at the University of Pisa concerning the application of an in-house developed coupling methodology between a modified version of RELAP5/Mod.3.3 and the ANSYS Fluent commercial CFD code to a pool system. Mono-dimensional codes, like RELAP5, are commonly used for thermal-hydraulic analysis of entire complex systems. Nevertheless, their one-dimensional feature represents a limit in the analysis of such problems where significant 3D phenomena are involved. On the other hand, CFD codes standalone are usually employed to simulate relatively small domains. The use of System Thermal-Hydraulic + CFD coupled calculations can overcome these issues, allowing the simulation of a complete system, but with a part of the domain reproduced with the CFD code.
In this work, the coupled calculation technique was used to simulate a PLOHS + LOF transient in the HLM experimental facility CIRCE (CIRCulation Experiment), located at the ENEA Brasimone research centre. The paper initially calls up the coupling procedure adopted, consisting in a “two-way” coupling. MATLAB software, used as external interface, manages the exchange of data between the system and the CFD code. The numerical method adopted for the coupling is the implicit scheme. Then, the main features of the CIRCE facility are briefly described, so are the two computational domains employed in this study. In particular, the CFD code was used to model the CIRCE pool (8 m high) and the Decay Heat Removal (DHR) heat exchanger. Due to the long duration of the transient simulated, a 2D axial-symmetric domain was chosen in order to reduce the computational time. The test section, placed inside the pool and consisting in a heat source and a heat sink, and the secondary side of the heat exchanger, were modeled with RELAP5. The use of the coupling tool allowed to set realistic boundary conditions in the calculation, more representative of the experimental ones. The main numerical results obtained from the PLOHS + LOF coupled calculation were compared with experimental data. Calculated LBE mass flow rates in the test section and in the DHR showed good agreement with experimental data. Some discrepancies with respect to the experimental trends were noticed for LBE temperatures; these should be related to some simplifications introduced in the model. Nevertheless, obtained outcomes represent a preliminary guideline for the improvement of the modeling for future works.
CFD Analysis of Thermal Mixing and Mass Flux Distribution in the PWR After MSLB
Authors: Xu Hong , Xu Yiban , Cao Liping , Sung Yixing , Kucukboyaci Vefa N. , Wright Richard F. 
Published: October 2017

During a postulated main steam line break (MSLB) event of a Pressurized Water Reactor (PWR) initiated at the Hot Zero Power (HZP) condition, increased heat removal from the broken steam generator (SG) on the secondary side that significantly reduces the coolant temperature on the primary side, and cold primary coolant enters the reactor vessel through the affected loop resulting in asymmetric temperature and mass flux distributions into the reactor core. A plant safety analysis under the MSLB condition needs to account for the thermal and mass flux asymmetry effects on the reactor core response due to the colder water flowing from the affected SG and the reactor coolant system (RCS) to reactor vessel.
High resolution computational fluid dynamics (CFD) methodology with ANSYS CFX (Version 16.1) software was applied to analyze the flow behaviors and thermal-hydraulic phenomena and to study the thermal mixing and asymmetry effects in the downcomer and lower-plenum of a typical Westinghouse design four-loop PWR under the MSLB conditions. Two scenarios were considered for the CFD simulation distinct by reactor coolant pump status:
(1) Low-flow case: without offsite power where the reactor core is cooled through natural circulation
(2) High-flow case: with offsite power available and the reactor coolant pumps in operation
The CFX CFD modeling and simulation were based on the reactor vessel boundary conditions from a system code transient simulation at the limiting time steps with respect to thermal margin of the fuel design. The geometric model included the vessel downcomer and the lower internals up to the reactor core inlet below the fuel assemblies. The results of CFD simulation show the different flow patterns and temperature distributions at the reactor core inlet for the low-flow case and for the high-flow case. Thermal asymmetric effect exists in both cases, but in the low-flow case, cold flow enters into core inlets at the opposite side of faulted loop located, and in the high-flow case cold flow enters into core inlets at the same side of faulted loop located. A mass flux asymmetric effect exists in both cases, but for the low-flow case, the core inlet mass flow distribution is more uniform than that for the high-flow case. The reactor core inlet distributions under the MSLB condition were further evaluated through comparisons with the results from the STAR-CCM+ (Version 10.04.01) CFD modeling and simulation. The evaluation showed that the simulation results are in good agreement with the STAR-CCM+ predictions and consistent with the phenomenon observed in an experiment published in open literature and site engineer judgment based on the available detected data.
Assessment of the Turbulent Prandtl Number Effect on Simulating Heat Transfer Characteristics of Supercritical Water Flow in Bare Tubes
Authors: Farah Amjad , Harvel Glenn , Pioro Igor 
Published: October 2017

Computational Fluid Dynamics (CFD) is a numerical approach to modelling fluids in multidimensional space using the Navier-Stokes equations and databases of fluid properties to arrive at a full simulation of a fluid dynamics and heat transfer system.
The turbulence models employed in CFD are a set of equations that determine the turbulence transport terms in the mean flow equations. They are based on hypotheses about the process of turbulence, and as such require empirical input in the form of constants or functions, in order to achieve closure. By introducing a set of empirical constants to a model, that model then becomes valid for certain flow conditions, or for a range of flows.
Of those constants, the turbulent Prandtl number appears in multiple equations; energy, momentum, turbulent kinetic energy, turbulent kinetic energy dissipation rate, etc. and the value it takes in each equation is different and chosen empirically to fit a wide range of flows in the subcritical region. The studies that attempt to find the effect of varying the turbulent Pr number on simulation results, often only mention one number; presumably the one that appears in the energy equation (although it is never explicitly explained). The rest of the constants are treated as universally acceptable for generalized flow and not tested for their effect on flow parameters.
A numerical study on heat transfer to supercritical water flowing in a vertical tube is carried out using the ANSYS FLUENT code and employing the Realizable k-ε (RKE) and the SST k-ε turbulence models. The 3-D mesh consists of a 1/8 slice (45° radially) of a bare tube. The study explored the effects of turbulent Pr numbers, and their variations, in order to understand their significance, and to build on previous knowledge to modify the turbulence models and achieve higher accuracy in simulating experimental conditions.
The numerical results of 3D flow and thermal distributions under normal and deteriorated heat transfer conditions are compared to experimental results. The distributions of temperature and turbulence levels are used to understand the underlying phenomena of the heat transfer deterioration in supercritical water flows.
Reducing the energy turbulent Pr number produced the most accurate prediction of the deterioration in heat transfer, by altering the production term due to buoyancy, which appears in the equations for turbulent kinetic energy as well as its dissipation rate. The buoyancy forces in upward flows act to reduce the turbulent shear stress, resulting in localized increase in wall temperatures.