bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of Assessment Method for a Self-Leveling Behavior of Debris Bed and Analyses of Experiments
Authors: Tagami Hirotaka , Cheng Songbai , Tobita Yoshiharu , Guo Liancheng , Zhang Bin , Morita Koji
Published: November 2014

When core melt occurs in severe accident in Sodium Cooled Fast Reactor (SFR), molten core material moves to the lower plenum in reactor vessel and fragmented by fuel coolant interaction. These fragmented particles, so called debris, accumulate on the structure surface to form debris bed. If the thickness of the debris bed exceeds the coolable thickness of the decay heat, boiling of sodium occurs inside the debris bed. It is found from past in-pile experiments that the sodium flow and boiling inside the debris bed caused by a decay heat planarize the debris bed to lower the debris bed thickness. This mechanism is called self-leveling of debris bed. In the accident sequence of SFR, when fuel debris locally accumulates beyond the coolable thickness, fuel debris remelts with decay heat and they cannot be retained in-vessel. However, it is expected that the debris bed thickness lowers the coolable thickness with self-leveling phenomenon and they can be safely retained in-vessel. This is why an appropriate assessment for self-leveling behavior is important for safety analysis of SFR with the object of safety cooling of fuel debris. Therefore, the object of this study is to develop new analytical methods to simulate unique phenomena in self-leveling behavior and implement it to SFR safety analysis code. The characteristic of self-leveling is that when the larger external forces caused by environmental fluids are larger than a threshold value, the debris bed is fluidized. The new methods are developed with assuming that the debris bed behaves as Bingham fluid from this feature. They are categorized into two main parts. The first part is particle interaction models to model the effect of particle-particle contacts and collisions. Particle pressure and particle viscosity related to particle-particle collisions and contacts, respectively, are applied to pressure and viscosity term in the particle momentum equation. The second part is a large deformation method, which simulates Bingham fluid characteristic of debris bed. This method numerically judges a onset of debris bed fluidization which depends on a shear stress strength. An experimental study of self-leveling behavior, in which the particle bed behavior driven by bubbles inflow from the bottom of bed in gas-solid-liquid three-phase flow was observed, is analyzed to validate the new methods. Simulation results well reproduced the transient changes of particle bed, whose elevation angle and form deformation becomes gradually small and obscure, respectively. Their dependencies on particle size and density are also well simulated with new methods. The assessment results show that these methods provide a basis to develop analytical methods of self-leveling behavior of debris bed in the safety assessment of SFRs.
Influence of Passive Safety Systems on a Cladding Temperature Condition of Fuel Rod Simulators
Authors: Melikhov Oleg I. , Elkin Ilya V. , Melikhov Vladimir I. , Nikonov Sergey M. , Pylev Sergey S. , Zaitsev Sergey I. , Zakutaev Maxim O. , Dorofeev Dmitry I.
Published: November 2014

In 2012 on the large-scale thermalphysic PSB-VVER facility modelling of three different type accidents is executed:

- Guillotine rupture of the pipeline on a reactor inlet;

- A small leak from “the cold” pipeline;

- Rupture of pressurizer surge line.

In experiments work of passive safety systems of new projects of NPP with RU VVER in the conditions of loss of all sources of an alternating current was modelled.

The purpose of experiments was research of influence of new passive safety systems on a temperature condition of a fuel cladding. On the basis of the received experimental data verification of the Russian system codes TRAP-KS, KORSAR/GP and SOCRAT is executed.

In paper the basic results of researches are presented.
Development of an Evaluation Methodology for the Fuel-Relocation Into the Coolant Plenum in the Core Disruptive Accident of Sodium-Cooled Fast Reactors
Authors: Kamiyama Kenji , Tobita Yoshiharu , Suzuki Tohru , Matsuba Ken-ichi
Published: November 2014

In the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs), fuel relocation out of the core region decrease the potential for severe power excursions caused by recriticality and control-rod guide tubes (CRGTs) provide an effective path for fuel-relocation. Therefore, a methodology for evaluating molten-fuel relocation through CRGTs is required in order to realistically evaluate sequences and consequences of CDAs. Since the liquid sodium exists at the coolant plenum, pressurizations and coolant void development associated with fuel-coolant interactions (FCIs) are considered to affect fuel-relocation. Therefore, the objective of the present study is to develop the methodology for evaluation of molten-fuel relocation into the coolant plenum with FCIs. In the present study, the SIMMER code which has been developed for CDA analyses was utilized as a technical basis since this code can treat multi-phase, multi-component fluid dynamics with phase-changes supposed to take place in the coolant plenum during fuel-relocation. The evaluation methodology was developed through validations of the SIMMER code using experimental data. A series of fundamental experiments were selected for model validations in which an alloy with low melting temperature and water were used as simulant materials for the fuel and the coolant respectively since the experiments were performed under a simulated CDA condition of SFRs in which a liquid-liquid direct contact was maintained between the melt and water contact surface, and the visual observation on FCI process was effective to validate models based on phenomenological considerations. The code was validated by two steps: In the first step, fundamental validations of melt-discharge into the coolant were performed, namely, momentum exchange functions of flowing-melt both to the wall of relocation-path and to the coolant were validated based on experimental data in which effects of FCIs on melt-discharge into the coolant were eliminated or negligible. In the second step, comprehensive validations of melt-ejection into the coolant were performed, namely, models which affect heat-exchange between the melt and the coolant were validated based on experimental data in which the melt was relocated into the coolant with FCIs. The second step validation required model improvements for suppression of melt-coolant interfacial area based on the results of visual observation in the experiments in order to reproduce the experimental results appropriately. Through the present model validations, the methodology to evaluate molten-fuel relocation into the coolant plenum with FCIs was successfully developed.
A Procedure for the Pre-Conceptual Design of Fast Reactors and Application to a Gas-Cooled Sub-Critical Transmuter
Authors: Fiorina Carlo , Mikityuk Konstantin , Křepel Jiři
Published: November 2014

A C++ procedure has been developed for the design and optimization of Fast Reactor (FR) cores. It couples the ERANOS based EQL3D procedure developed at PSI for FR equilibrium fuel cycle analysis with a dedicated MATLAB script that evaluates the thermal-hydraulic characteristics of the reactor core. It is conceived to investigate reactors with both standard pins and annular pins. The procedure accepts as input the physical properties of the system, as well as a set of target core parameters presently consisting of core power, maximum fuel burnup, multiplication factor, inner pin diameter (for annular pins) or maximum pressure losses (for standard pins), and core height. It gives as a result a core design fulfilling these design objectives and meeting the constraints on maximum fuel and clad temperatures. In case of annular pins, it also equalizes the temperature rise inside and outside of the core average pin. The procedure considers the possibility of two-zone cores and adjusts the fuel composition in the two zones to achieve an optimal radial power distribution. Finally, it can evaluate safety parameters and fuel cycle characteristics both at beginning-of-life and at equilibrium. As a test case, the procedure has been used for the pre-conceptual design of a sub-critical Gas Fast Reactor core employing inert-matrix sphere-pac fuel and annular pins with SiC cladding.
Experimental Research on Pulsating Two-Phase Flow Pressure Drop Characteristics in Horizontal Rectangular Channel
Authors: Zhang Siqi , Gao Puzhen
Published: November 2014

In spite of most previous studies since 1970, the theory of pulsating pipe flows supported by experimental investigations has not yet completed in comparison with the well-defined theory of steady pipe flows. Therefore, it seems that there is much to be done about experimental research in this field. In order to determine the resistance characteristics of two-phase flow under pulsatile conditions, an experimental investigation on two-phase flow with periodically fluctuating flow rates in a narrow rectangular channel is carried out. A frequency inverter is used to obtain experimental conditions with different fluctuating frequencies, amplitudes and mean values of water mass flow rate. After obtaining experimental results, comparisons between experimental frictional pressure drop values and theoretical calculations have been done. Two-phase flow on pulsating conditions is far more complicated than that on steady conditions because pulsating flow is composed of two parts: a steady component and a superimposed periodical time varying component called oscillation. In this paper, the influence of different fluctuating frequencies, amplitudes and mean values of liquid and gas mass flow rate on two-phase flow pressure drop characteristics is also discussed.

The results show that the total pressure drop and water mass flow rate change with the same fluctuating period except for a phase difference. The phase lag also changes with the fluctuating frequencies and amplitude. The accelerating pressure drop changes dramatically in a fluctuating period, especially at the end of acceleration. Also, the time when the acceleration pressure drop has its maximum value lags the time when the acceleration reaches its peak, mainly because of the inertial of the fluid.
Molten Salt Reactor With Simplified Fuel Recycling and Delayed Carrier Salt Cleaning
Authors: Křepel Jiři , Bykov Valentyn , Mikityuk Konstantin , Hombourger Boris , Fiorina Carlo , Pautz Andreas
Published: November 2014

The Molten Salt Reactor (MSR) represents an old concept, but its properties are qualifying it for the advanced utilization: inherent safety, excellent neutron economy, possibility of continuous or batch reprocessing without fuel fabrication. The aim of this paper is to characterize the MSR unique fuel cycle advantages in different neutron spectra using the results of ERANOS-based EQL3D and ECCO-MATLAB based EQL0D procedures. It also focuses on the low production of higher actinides in the Th-U cycle and based on the results, it proposes a simplified in situ recycling of the fuel and the delayed ex situ carrier salt cleaning or direct disposal by vitrification.
Safety Design Features for Defense-in-Depth Against Beyond Design Basis Accidents for Cernavoda NPP: Units 3 and 4
Authors: Dinh Ngoc-Boi , Ford Steven , Lekakh Boris
Published: November 2014

The Cernavoda Units 3&4 (C3&4) CANDU 6 design is the upgraded Cernavoda 2 CANDU 6 design that meets the applicable Romanian regulatory requirements (NSN-2, NSN-8) and the WENRA Reactor Safety Reference Levels requirements. C3&4 build on the proven high performance design and the defence-in-depth features of CANDU 6 units and incorporate further improvements to safety and operational performance.
Estimation of Ex-Vessel Steam Explosion Pressure Loads for AP1000 Nuclear Power Plant
Authors: Li Zhigang , Li Jinghao , Lin Meng , Yang Yanhua
Published: November 2014

An ex-vessel steam explosion is a fuel coolant interaction process which may occur when the reactor vessel fails and the molten core pours into the water in the reactor cavity during a severe reactor accident. A strong enough steam explosion in a nuclear power plant could endanger the containment integrity and lead to a direct release of radioactive material to the environment. In this article, a nuclear island geometrical model of AP1000 nuclear power plant was established and different scenarios of ex-vessel steam explosions in AP1000 NPP were simulated by MC3D code. Since the initial parameters with large quantity of uncertainties under accident condition may have important effects on the steam explosion, some initial parameters study were performed by varying the location of the melt release (75°,45°,30°,0°), the cavity water subcooling, the triggering time for explosion calculations, the melt temperature and the break size. Results indicate that the higher the melt temperature, the longer the triggered time and the lower the coolant temperature would lead to the more severe steam explosion. Besides, when the angle of break reaches 45 degree and the diameter of the break is 0.5m, the steam explosion causes the largest damage.
A Study of Heat Transfer Calculation Method of Reactor Vessel Metallic Insulation
Authors: Dapeng Yan , Luo Ying
Published: November 2014

Metallic insulation is commonly used in reactor vessel because of its resistance to radiation and corrosion. Since the main mode of heat loss of reactor vessel is thermal radiation, the ability to prevent radiation heat transfer is important for metallic insulation. But the thermal conductivity of metallic insulation is difficult to calculate owing to their complex geometry. This article uses FLUENT 14.0 to obtain the important parameter “view factor”, and then develops a computational model of effective conductivity of metallic insulation. Heat transfer test of metallic insulation was done, and the numerical simulation of metallic insulation was also performed. Based on results of test and simulation, the computational model is modified. The modified model can fit the test result better. Based on the modified model, the effective conductivity of metallic insulation increases with the increase of temperature of hot side and cold side, among which the temperature of hot side influences more. And when the temperature is high, the effective conductivity increases much faster.
Prediction of Laminar to Turbulent Transition in 90° Curved Duct Based on Energy Gradient Theory
Authors: Fu Xuekuan , Yuan Hongsheng , Chen Chong , Chen Hanying , Song Yulin , Tan Sichao
Published: November 2014

Study on laminar to turbulent transition in a 90° curved duct is carried out. Professional mesh generation tool ICEM is used to create geometry model and generate mesh, then the mesh is imported to the fluid dynamics software CFX to simulate three-dimensional flow. The dimensionless parameter K is calculated in the duct with velocity and pressure distribution obtained by numerical simulation. According to energy gradient theory, K is the ratio of energy gradient in the transverse direction to that in the streamwise direction, which can be used to measure stability of the flow field. It is found that fluid is most likely to lose its stability where the maximum K is, and laminar to turbulent transition occurs when K reaches a certain value, corresponding to the critical Reynolds number. The instability and transition characteristics predicted by energy gradient method agree well with former researchers’ results.
Numerical Analysis of the Weld Joint Resulting Influence on the Pressurizer Surge Line Stress Distribution
Authors: Yang Yang , Xia Hong , Zeng Xian , Yu Dali
Published: November 2014

The thermal stratification phenomenon widely exists in the pressurizer surge line (PSL) during the steady-state conditions and transient conditions, which could result an unexpected stress distribution. Especially, the maximum stress range is most likely to appear nearby the weld joint area on the PSL. This paper mainly focuses on the pressurizer surge line stress distribution variation caused by the weld joint. The temperature distribution and stress distribution of the surge line are calculated by using the finite element method (FEM) and finite volume method (FVM). Steady-state operating conditions are considered. The conclusion of this work is that the stress distribution is quite different from the traditional stress profile merely caused by the upper and lower temperature stratification. Moreover, the stress is much higher in the interact-area between PSL matrix and the weld joint than in any other areas.
Development of High Accurate Flow Nozzle for Feedwater Flowrate Measurement in Nuclear Power Plant
Authors: Furuichi Noriyuki , Cheong KarHooi , Terao Yoshiya , Nakao Shinichi , Fujita Keiji , Shibuya Kazuo
Published: November 2014

The high accurate throat tap flow nozzle with four different diameter taps is developed and its discharge coefficients are measured in the Reynolds number range from 1.5×106 to 1.4×107 using the high Reynolds calibration facility of AIST,NMIJ. The discharge coefficient of a throat tap nozzle extrapolated according to ASME PTC 6 are confirmed to deviate 0.37% at Red=1.4×107 from the experimental results. The high accurate flow nozzle developed can reduce this extrapolation error of the discharge coefficient to high Reynolds numbers by using the equations of discharge coefficients, which is determined as a function of Reynolds number and tap diameter based on the experimental results of four different diameter taps. The error of extrapolated discharge coefficient using the derived equations is estimated to be less than 0.1% at Red=1.4×107. The present results show that the throat tap flow nozzle developed is expected to work as a high accurate flowmeter even under the extrapolation of the discharge coefficient toward high Reynolds numbers.
Analysis of a Loss of Heat Removal Accident in a PWR Spent Fuel Pool
Authors: Wu Xiaoli , Zhang Yapei , Tian Wenxi , Su Guanghui , Qiu Suizheng
Published: November 2014

The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents a study on the consequences of loss of heat removal accident in the spent fuel pool of a typical pressurized water reactor using the Modular Accident Analysis Program (MAAP5) code. Analysis of uncompensated loss of water due to the loss of heat removal with initial pool water level of 12.2 m (designated as a reference case) has been performed. The analyses cover a broad spectrum of severe accident in the spent fuel pool. Those consequences such as overheating of uncovered fuel assemblies, oxidation of zirconium and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission product are also analyzed in this paper. Furthermore, as important mitigation measures, the effects of makeup water in SFP on the accident progressions have also been investigated based on the events of spent fuels uncovery. The results showed that spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate higher than evaporation rate defined in the reference case. Although spent fuel assemblies partly exposed due to a mass flow rate of makeup water smaller than the average evaporation rate, continuous steam cooling and radiation heat transfer might maintain the spent fuels coolability as the actual evaporation was balanced by the makeup in a period of time of the order of several days. However, larger makeup rate should be guaranteed to ensure long-term safety of SFP.
Concept of Standard System Based on China Current Nuclear Power Construction
Authors: Yan Yuelan
Published: November 2014

Since introduce different technical routes, during decades of nuclear power development in our country, the French RCC series standards, American ASME standards and Russian standards are adopted, which led to the current various standards exist in their own way. To promote the building of nuclear power standards system in China, in the year of 2012, important research subject “the research on the standard system of advanced nuclear power in China” has been carried out and subject “nuclear power construction and commissioning” is one of it.. By digestion and absorption of four oversea AP1000 units of Sanmen nuclear power plant in Zhejiang province and Haiyang nuclear power plant in Shandong province, the building of standard system during nuclear power construction suitable to our national condition is studied, including the system frame and composition standards, building standard system method during construction, namely through research and example to present what kind of standard system is suitable for China standard system during construction, and what kind of method or design is used to obtain and maintain such system. The thesis is to promote the subject research methods based on examples to build China’s nuclear power standard system.
The Research of Suitable Collimator Used in Gamma Scanning System
Authors: Qian Nan , Wang Dezhong , Gu Weiguo , Ma Yuanwei
Published: November 2014

Collimator is an important equipment in gamma scanning system. Visual field and measurement uniformity are important metric on properties of collimator. This paper studied the influence of the collimator shape and size on the visual field and measurement uniformity. The small and large square windows were not the optimal. Comparing those properties, a special shape collimator was introduced in this research. By contrast, its properties were improved. Then, we made several groups of experiments with the orthogonal experimental design, and discussed the effect of size parameters on the visual field and measurement uniformity. The length of the collimator is the most important factor. The influence of size parameters is the same on measurement uniformity. Then, we obtained the optimal size parameters.
Elaboration of the System Based Code Concept — Activities in JSME and ASME: (1) Overview
Authors: Asayama Tai , Miyagawa Takayuki , Dozaki Koji , Kamishima Yoshio , Hayashi Masaaki , Machida Hideo
Published: November 2014

This paper is the first one of the series of four papers that describe ongoing activities in the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) on the elaboration of the System Based Code (SBC) concept aiming at its application to nuclear structural codes and standards. This paper includes a brief introduction to the SBC concept and the technical features of structural evaluation methodologies that are being developed for use in the SBC concept; Load and Resistance Factor Design based reliability assessment methods and the JSME guidelines for reliability evaluation of static components of sodium-cooled fast reactors. Also described is the ongoing collaboration of JSME and ASME at the Joint Task Group for System Based Code established in the ASME Boiler and Pressure Vessel Code Committee which is developing alternative rules for ASME B&PV Code Section XI Division 3, inservice inspection requirements for liquid metal reactor components.
Two-Phase Flow Induced Force Fluctuations on Pipe Bend
Authors: Miwa Shuichiro , Liu Yang , Hibiki Takashi , Ishii Mamoru , Kondo Yoshiyuki , Morita Hideyuki , Tanimoto Koichi
Published: November 2014

In this study, fluctuating force induced by both upward and horizontal gas-liquid two-phase flow on 90 degree pipe bend at atmospheric condition was investigated. First, the database comprised of dynamic force signals and two-phase flow parameters such as volumetric fluxes, area averaged void fraction and pressure fluctuations covering entire two-phase flow regimes was developed for both flow orientations. Then, study was conducted to develop a model which is capable of predicting the force fluctuation frequency and magnitudes particularly for the slug flow regime. The model was fundamentally developed from the local instantaneous two-fluid model which was applied to the control volume around the elbow test section. Main contribution of the force fluctuation of two-phase flow is from the momentum and pressure fluctuations for most of the flow regimes. For slug flow regime, however, water-hammer like impact was produced by the collision of liquid slug against the structure surface. In order to consider that effect, the liquid slug impact force model was developed. The model utilizes two-group interfacial area concentration correlation to treat the flow regime transition without an abrupt discontinuity. It was found that the newly developed model is capable of predicting two-phase flow induced force fluctuation and dominant frequency range with satisfactory accuracy for flow regimes up to churn-turbulent.
COUPLE, A Time-Dependent Coupled Neutronics and Thermal-Hydraulics Code, and its Application to MSFR
Authors: Zhang Dalin , Zhai Zhi-Gang , Rineiski Andrei , Guo Zhangpeng , Wang Chenglong , Xiao Yao , Qiu Suizheng
Published: November 2014

Molten salt reactor (MSR) using liquid fuel is one of the Generation-IV candidate reactors. Its liquid fuel characteristics are fundamentally different from those of the conventional solid-fuel reactors, especially the much stronger neutronics and thermal hydraulics coupling is drawing significant attention. In this study, the fundamental thermal hydraulic model, neutronic model, and some auxiliary models were established for the liquid-fuel reactors, and a time-dependent coupled neutronics and thermal hydraulics code named COUPLE was developed to solve the mathematic models by the numerical method. After the code was verified, it was applied to the molten salt fast reactor (MSFR) to perform the steady state calculation. The distributions of the neutron fluxes, delayed neutron precursors, velocity, and temperature were obtained and presented. The results show that the liquid fuel flow affects the delayed neutron precursors significantly, while slightly influences the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank, which leads to the maximal temperature about 1100 K at the centre of the vortex. The results can provide some useful information for the reactor optimization.
Assessment of APR-1400 Emergency Core Cooling System Performance for Design Basis LOCA Redefinition
Authors: Kang Dong Gu , Ahn Seung-Hoon , Chang Soon Heung , Huh Byung-Gil , Bang Young-Seok , Seul Kwang-Won
Published: November 2014

As a part of the efforts to develop the risk-informed regulation, alternative rulemaking of 10CFR50.46 is underway. In the rule, USNRC divided the current spectrum of LOCA break sizes into two regions, by determining a transition break size (TBS), and the LOCAs for any breaks larger than TBS would be regarded as beyond design basis accident (BDBA). A combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of BDBAs. The performance of the APR-1400 emergency core cooling system (ECCS) performance was assessed against large break LOCA applying CDPP. It was confirmed that current APR-1400 ECCS design has capability to mitigate BDB LOCA by analyzing ECCS cooling performance for BDB LOCA. The proposed CDPP was also applied to design changes of the emergency diesel generator (EDG) start time extension and power uprates with simplified assumption that the probabilistic safety assessment (PSA) data are still valid. By assumptions and considerations, the CDPP to assess ECCS performance for plant design modification was reduced to calculating conditional exceedance probability (CEP) of one sequence and comparing allowable value. The allowable CEP was used to determine whether the design change is acceptable or not, and discussions were made for acceptable nuclear power plant changes.
A Study on Performance Characteristics for Closed Brayton Cycle in Inventory Regulation
Authors: Jiang Huijing , Yang Xiaoyong , Ding Ming , Wang Jie
Published: November 2014

High temperature gas-cooled reactor combined with gas turbine cycle can make full use of the high outlet helium temperature of reactor. Compared to the Open Brayton Cycle, inventory regulation is a new regulation method for Closed Brayton Cycle, which can control the output power of the cycle through the change of pressure level and mass flow of the system. Taking the 10MW high temperature gas-cooled test reactor coupled with helium-turbine cycle (HTR-10GT) for example, a more practice-oriented resistance model is established in this paper for the analysis of influence of resistance characteristic on key components and closed Brayton cycle. When helium inventory drops from 100% to 30% of rated value, the mass flow rate of helium decreases, so the local pressure loss coefficient keeps constant and the pressure loss coefficient increases. As a result, the efficiency of closed cycle decreased about 15%. The performance characteristics of the turbomachinery and heat exchangers are also discussed in detail as key components in the power conversation unit (PCU).
Transient Validation of Areva’s New Arcadia™ Code System
Authors: Kuch Sebastian , Leberig Mario , Brock Richard , Reiterer Florian , Riedmann Michael , Rooks Kenneth
Published: November 2014

AREVA has developed a new leading edge code suite to meet the challenges arising from increasing expectations in nuclear power plant availability and fuel performance while satisfying stricter safety requirements. ARCADIA™ [1] is an advanced 3D coupled neutronics/thermal-hydraulics/thermal-mechanics code system for Light Water Reactor (LWR) fuel assembly and core design calculations as well as safety analysis, using a new software architecture allowing for nodal and pin-by-pin calculation capability. ARCADIA™ was licensed by the US Nuclear Regulatory Commission (NRC) for applications for PWR UO2 cores in 2013. It is on the way to be licensed in other countries for AREVA customers.

ARCADIA™ contains the steady-state and transient core-simulator ARTEMIS™ [2] for core design and coupled transient safety analysis. ARTEMIS™ can be used in a coupled mode with S-RELAP5 and CATHARE 2 to allow fully coupled transient analysis, combining the sophisticated 3D core model of ARTEMIS™ with the proven system thermal-hydraulics of S-RELAP5 and CATHARE 2 including a detailed simulation of the Instrumentation and Control (I&C). This allows simulating complex transients affecting the core as well as the primary and secondary side including I&C signals and responses.

For the validation of ARTEMIS™ a comprehensive set of validation cases was selected, including international benchmarks and measurements covering various classes of transients. These cases include a ‘Load Rejection to station service’ event at a German 1300 MWe plant, where a wide range of system and core parameters was measured that allow the validation of the fully coupled code system. Another validation case is a nodal recalculation of the core behavior during the pump shaft break transient that occurred in the Gösgen nuclear power plant in 1985 [3].

The paper will provide representative example results for the abovementioned validation cases.
HELENA: A Heavy Liquid Metal Multi-Purpose Loop for Thermal-Hydraulics, Corrosion and Component Test
Authors: Di Piazza I. , Tarantino M. , Agostini P. , Gaggini P.
Published: November 2014

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved on the HLM technology development.

Currently ENEA has implemented large competencies and capabilities in the field of HLM thermal-hydraulic, coolant technology, material for high temperature applications, corrosion and material protection. In this frame, the HELENA facility is well instrumented and it represents a loop working in pure lead for experiments in the field of corrosion for LFR structural materials, component test, and thermal-hydraulic investigations. The components and the process of the facility has been also depicted in some details. The scheduled working activities has been described and the main steps focused.

A prototypical mechanical pump has been designed and manufactured to properly work in pure lead at high temperatures. The long-run test on the component will be isothermal at 400°C with low oxygen content <10−8 % in weight; the oxygen content will be monitored continuously. The pump will be gradually driven to the reference mass flow rate 35 kg/s and this mass flow rate will be maintained for 1500 h, i.e. 2 months about. After this, the pump is stopped and the loop is drained. Then, the pump impeller is disassembled by the body and it is analyzed for corrosion. Then, a test on the ball valves is carried out for a few months. At the end of 2014, the facility is upgraded with the insertion of the FPS in the heating section and with the secondary side. A 19-pin wire-spaced Fuel Pin bundle Simulator (FPS) is installed to measure clad temperature and heat transfer coefficients in different conditions in the different ranks of sub-channels of the MYRRHA bundle. Therefore a test matrix on the forced convection condition in the wire-spaced bundle will be carried out.

A 7 tube/shell-and-tube Heat Exchanger couples the primary lead loop with the secondary side with water in pressure at 100 bar. The tube-in-tube technology with lead tube side, water shell side, steel powder in the gap is adopted. Bubble tubes with flowing Argon are adopted to measure pressure losses in the different branches of the loop. Several thermocouples monitor the loop in different points. An ancillary gas system ensures the cover gas.

The paper reports the description of the experiments, the proposed test matrix and description the technological solution adopted for the HELENA implementation.
Analysis of the Control Rod Drop Accident (CRDA) for Lungmen ABWR
Authors: Ho Ai-Ling , Wang Jong-Rong , Lin Hao-Tzu , Shih Chunkuan
Published: November 2014

The object of this paper is to understand the realistic behavior in Lungmen ABWR (Advanced Boiling Water Reactor) during a control rod drop accident (CRDA) transient. The CRDA transient would lead the reactor through an extremely fast and localized power excursion, requiring an accurate core modeling. The CRDA analysis for Lungmen ABWR was performed by coupling the 3D neutron kinetic code, PARCS, and two-phase thermal-hydraulic (T-H) code, TRACE. After TRACE/PARCS coupling calculation, the output data from TRACE/PARCS would be inputted into FRAPTRAN code as a function of time-dependent fuel rod power and coolant boundary conditions to calculate the fuel damage.

The CRDA analysis for Lungmen ABWR was performed for two conditions: a) case1: hot-full-power (HFP) at beginning of cycle (BOC); b) case2: hot-zero-power (HZP) at BOC. Under these conditions, the damage mechanisms of fuel rod are: 1) cladding ballooning and burst; 2) embrittlement and failure by high-temperature oxidation; 3) melting of cladding and/or fuel pellets. And the relevant quantities for fuel performance are the maximum fuel enthalpy and the melting temperatures of cladding and fuel pellet.

The results of CRDA analysis show that a) case1: no fuel failure occurs under HFP condition at BOC; b) case2: the fuel rod nearby the dropped control rod failed under HZP condition at BOC, and the FRAPPTRAN data exposes that the main reason of rod failure is the cladding high temperature.
Awareness of Danger: A Factor May Improve Public Acceptance to Nuclear if Included in Education
Authors: Liu Mingye , Xia Hong
Published: November 2014

In this paper, facts are traditionally considered as negative in nuclear related education, e.g. the invisible nuclear radiation, harmful release, even nuclear accidents, are discussed. This paper puts that the factors may let the public be against to nuclear are possible to be converted and improve public acceptance to nuclear industry. The premise is: awareness of those factors are included in nuclear education as a highlighted part, unlike the way they were treated before: never be mentioned or even be hidden on purpose. Nuclear education used to merely show the “perfection” of nuclear power, such as it is clear, green and effective, which is absolutely correct. However, people know that there is no silver bullet. The disadvantages of nuclear are almost never included in nuclear education, not mention the accidents or solutions to nuclear emergency. This type of education helps little to improve public acceptance to nuclear, what is worse, it prevents the public from seeing a whole picture of nuclear and nuclear industry; the public may be puzzled and tend to reject nuclear, because that they don’t have proper awareness of danger, and tend to magnify it indefinitely. This paper focuses on how the information and knowledge of nuclear radiation, NPP accident, and other “dark” facts may affect the impression of nuclear to the public. This paper also discusses the parts are necessary to be included in nuclear education, such as radiation protection, contingency plan and factors mentioned above. Additionally, this paper proves that if dealt correctly, the facts to be treated negative are turned to have a miraculous effect on improving public acceptance to nuclear.
Burnable Absorber Selection With UWB1 Depletion Code
Authors: Lovecký Martin , Prehradný Jan , Škoda Radek
Published: November 2014

Research of nuclear reactor fuel depletion aims at development and introduction of advanced types of burnable absorbers (BA) applied within nuclear fuel. BAs compensate for the initial reactivity excess and consequently may allow for lower power peaking factors and longer fuel cycles with higher fuel enrichments.

Modern computer codes for nuclear fuel depletion calculation require a substantial amount of computational time. Therefore, any parametric calculations for BA selection need to be carried out only with a fast depletion code. The main purpose of the newly developed UWB1 code is a rapid calculation of nuclear fuel depletion, which is achieved by the approximations in the equations describing transport part of fuel depletion. Microscopic cross sections are assumed to be constant through depletion calculation steps.

The paper describes the first step of analysis using a new version of UWB1, that was accomplished with the assumption of uniformly distributed BA in fuel. BA elements, nuclides and nuclide mixtures were compared and their performance was consequently evaluated based on multiplication coefficient behavior during depletion.
Numerical Analysis on Flow Inversion in a Uniformly Heated Thin Rectangular Channel Array
Authors: Park Jonghark , Park Suki , Jo Daeseong , Chae Heetaek , Lee Byungcheol
Published: November 2014

Most open-pool type research reactors, which are cooled by forced convection, have a downward flow for several advantages. Downward flow can minimize pool surface radiation, and makes a fuel assembly easy to be fixed on the grid plate of a reactor core. After LOFA (Loss of Flow Accident) or even normal shutdown, a reactor core has to go through a flow inversion from downward to upward owing to a natural circulation. As the pump is turned off, coolant flow in the core becomes slow and finally stagnant upon a flow inversion occurrence. During the flow inversion, it is a concern that the onset of nucleate boiling (ONB) or a departure from nucleate boiling (DNB) may occur. Recently, there have been several studies on predicting the flow inversion and temperature of a fuel plate to find whether an ONB occurs, using one-dimensional thermal hydraulic analysis codes and computational fluid dynamics (CFD) codes. These studies have concentrated on only a single channel but they did not consider other important effects such as the piping, valves, and pool. In the present study, a numerical simulation using a CFD code was carried out to determine the flow inversion phenomenon. The CFD model considers 21 thin plates of an in-line array which release heat uniformly. The 21 plates make 22 thin channels of which the inlets are connected to a large pool. The outlets of the thin channels are connected to the large pool through a flap valve. A CFD simulation was carried out during 80 seconds after the reactor shutdown. It shows the behaviors of a flow through the flap valve as well as the flow inversion in the channels. Another numerical simulation using the RELAP5 code was conducted to compare with the CFD simulation results. In the comparison, CFD predictions such as coolant temperature variation, heating plate temperature variation, and flow rate through the flap valve, show similar results with those of the RELAP5 analyses.
Burnable Absorber Comparison Between VVER, PWR and SFR With UWB1 and SERPENT Codes
Authors: Prehradný Jan , Lovecký Martin , Škoda Radek
Published: November 2014

Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers (BA) in nuclear fuel. BAs compensate for the initial reactivity excess and consequently allow for lower power peaking factors and longer fuel cycles with higher fuel enrichments.

The paper describes the comparison of selected BA elements (Gd, Eu, Er) for 3 different nuclear fuel types (VVER, PWR and SFR). The effect of harder VVER spectra and the effect of thermal vs fast neutron spectra is evaluated and commented. Uniformly distributed BAs in the fuel were assumed.

Comparison calculations were performed with the newly developed UWB1 fast fuel depletion code and with the SERPENT Monte Carlo code. The comparison of UWB1 and SERPENT codes is included in the analysis. Next steps of improving the UWB1 code are suggested.
Review on Water Radiolysis in the Fukushima Daiichi Accident: Potential Cause of Hydrogen Generation and Explosion
Authors: Saji Genn
Published: November 2014

Although the water radiolysis, decomposition of water by radiation, is a well-known phenomenon the exact mechanism is not well characterized especially for severe accidents. The author first reviewed the water radiolysis phenomena in LWRs during normal operation to severe accidents (e.g., TMI- and Chernobyl accidents) and performed a scoping estimation of the amount of radiological hydrogen generation, accumulation and release for the Fukushima Daiichi accident. The estimation incorporates the decay heat curve after a reactor trip combined with G-values. As much as 450 cubic meters-STP of accumulated hydrogen gas is estimated to be located inside the PCV just prior to the hydrogen explosion which occurred a day after the reactor trip in Unit 1.

When a set of radiological chain reactions are incorporated the resultant reverse reactions substantially reduce the hydrogen generation, even when removal of molecular products (i.e., oxygen and hydrogen) is assumed stripped rapidly from boiling water through bubbles. Even in the most favorable configuration a typical amount of hydrogen gas reduces to only several tens of cubic meters.

Finally, the author tested a new mechanism, “radiation-induced electrolysis,” which had been applied to his corrosion studies for last several years. His theory has been verified with the published in-pile test data, although he has never tried to apply this to his severe accident study. The predicted results indicated that the total inventory of hydrogen gas inside RPV may reach as much as 1000 cubic meters in just 3 hours during the SBO due to a high decay heat soon after the reactor trip through this process.
Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor
Authors: Cho Jaehyun , Shin Yong-Hoon , Hwang Il Soon
Published: November 2014

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.
Limit Load Analysis of Pipe Bend With Extrados Wall Thinning
Authors: Kim TaeRyong , Oh ChangKyun
Published: November 2014

Since pipe bend has a characteristic that extrados becomes thinner and intrados thicker after fabrication process, it can be expected to be vulnerable to extrados wall thinning due to corrosion or erosion during its operation. In this paper, limit loads of pipe bend with the thinning are computed under the loading conditions of internal pressure and bending moment. Several case studies with varying geometries and wall thinning shapes are presented. The difference in the limit loads behavior between pipe bend and welded elbow is also reviewed. The calculated plastic limit loads of pipe bend are compared with other research results for the welded elbow. The results show that pipe bend can be applied to safety-related piping systems as far as the internal pressure and bending moment only are considered.