bluuebunny/crossref_metadata_2025 huggingface dataset:
Influence of RCS Depressurization Strategy on Hydrogen Risk
Authors: Li Yabing , Tong Lili , Cao Xuewu
Published: November 2014

For advanced passive PWR, reactor coolant system (RCS) depressurization through automatic depressurization system (ADS) is an important measurement to avoid high-pressure melt ejection and direct containment heating. It allows injection from passive core cooling system and the implement of in-vessel retention. However, it has negative impact that hydrogen in the RCS can be released to the containment together with coolant, which may lead to hydrogen burning or even explosion in the containment. Therefore, this paper analyzes the RCS depressurization strategy during severe accident, and evaluates its negative impact. Severe accident sequences induced by station black out (SBO) was selected and analyzed with integral severe accident analysis code as a typical high pressure core melt accident scenario. Different depressurization strategies with ADS system were discussed based on Severe Accident Management Guideline (SAMG.) ADS valves were manually opened at a core exit temperature of 923 K with 20min delay for operator reaction. Both depressurization effect and hydrogen risk were evaluated for different strategies. Hydrogen distribution was calculated, which was used to determine the combustion mode in different compartments. Result shows all three strategies analyzed in this paper can depressurize the RCS effectively. And opening the ADS stage 1–3 valves causes rapidly increase of the hydrogen concentration in the in-containment refueling water storage tank (IRWST) compartment and may lead to hydrogen denotation. However, hydrogen can be well dispersed in the loop compartment with intentional open of ADS stage 4 valves to RCS depressurization. Therefore, suggestions are proposed for SAMG: implement RCS depressurization strategy with stage 4 ADS instead of ADS stage 1–3.
Hydraulic Lift-Off Issues for Application of High Performance Annular Fuels in Pressurized Water Reactors
Authors: Zhao Jiyun , Gui Jingkang
Published: November 2014

In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17×17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12×12 and 13×13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13×13 and 12×12 arrays) are several times larger than that of the referenced solid fuel (17×17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed.
Safety Feature of EU-ABWR for Fukushima Accident
Authors: Kamei Kazuhiro , Kuroda Masatomo , Kojima Yoshihiro , Yano Kazuki
Published: November 2014

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash, severe accident mitigation systems, the N+2 principle in safety systems, the diversity principle and a large output of 1600 MWe. These features enable EU-ABWR’s design objectives and principles to be consistent with the safety requirements of Western European Nuclear Regulators’ Association (WENRA) and Finnish YVL guides.

By designing safety features of the EU-ABWR based on Defense in Depth Principle, EU-ABWR has a capability to mitigate Design Extension Conditions like Fukushima Accident by using a combination of passive and active systems. Furthermore, a severe accident is mitigated passively without containment venting for at least 72 hours.
A Time-Dependent Solver for Coupled Neutron-Transport Thermal-Mechanics Calculations and Simulation of a Godiva Prompt-Critical Burst
Authors: Fiorina Carlo , Aufiero Manuele , Pelloni Sandro , Mikityuk Konstantin
Published: November 2014

The present paper describes a first step taken at the Paul Scherrer Institut in the development of a new multi-physics platform for reactor analysis. Such platform is based on the finite-volume software OpenFOAM and aims at a tightly coupled description of neutron transport, thermal mechanics and fluid dynamics. For this purpose, a steady-state 3-D discrete ordinates/thermal-mechanics solver was first developed in collaboration with the Politecnico di Milano. The present work briefly discusses such solver and its preliminary validation, which will be described in detail in parallel publications. It then focuses on its extension to time-dependent simulations. The solver is first tested by simulating different step-wise reactivity insertions in a critical configuration constituted by an infinite slab of highly enriched uranium. Subsequently, a super-prompt-critical power burst in the Godiva reactor has been simulated. Godiva was a spherical assembly of highly enriched uranium built and operated at the Los Alamos National Laboratory (US) during the Fifties. A prompt-critical transient in such system configures as a quick power excursion (up to ∼10 GW), which causes a temperature rise, and a subsequent reactivity reduction via expansion of the sphere. The overall transient lasts for few fractions of a millisecond. Results obtained with the newly developed model have been compared to experimental results, showing a relatively good agreement.
Compatibility of Ferritic/Martensitic Steel T91 With Flowing LBE Containing 10−7 Mass% Dissolved Oxygen at 450 and 550°C
Authors: Tsisar Valentyn , Schroer Carsten , Wedemeyer Olaf , Skrypnik Aleksandr , Konys Jürgen
Published: November 2014

Corrosion behavior of ferritic/martensitic (F/M) steel T91 with 9 % nominal chromium content was investigated in flowing (2 m/s) Pb-Bi eutectic (LBE) with 10−7 mass% dissolved oxygen at 450 and 550°C for up to 8766 and 2011 h, respectively. The corrosion process and material loss were analyzed. The steel generally shows oxidation consisting of the formation of an Fe-Cr spinel layer. Oxidation resulted in metal recession which does not exceed ∼ 10 μm after 8766 h at 450°C and ∼15 μm after 2011 h at 550°C. Local liquid-metal attack in the form of pits occurs at both temperatures. At 450°C, remarkable liquid-metal attack is observed after 5015 h, while, at 550°C, it is found already after 1007 h. The maximum depth of local attack reaches 960 μm after 8766 h at 450°C, and 190 μm after 1007 h at 550°C.

The obtained results are compared with previous experiments at similar testing conditions, except for a higher oxygen concentration of 10−6 mass%.
Validation Status of MAAP5 Core Melt Progression Model
Authors: Paik Chan Y. , McMinn Paul , Henry Christopher , Luangdilok Wison
Published: November 2014

The Modular Accident Analysis Program (MAAP) is a computer code that is used for integrated severe accident analysis. The latest MAAP5 version was validated against the PHEBUS-PF FPT0 and FPT1 tests performed at CEA/IPSN, PBF-SFD Test 1–4 performed at INEL, and QUENCH Test 06 performed at FZK. PHEBUS FPT0, PHEBUS FPT1, and PBF-SFD Test 1–4 are in-pile experiments where a test bundle was housed in the center of a reactor. Comparisons were focused on fuel and shroud temperature histories, and hydrogen generation histories. For the PHEBUS tests, primary system and containment responses were also compared. In general, fuel and shroud temperatures are well predicted by MAAP5. Overall hydrogen mass generated is also well predicted except that MAAP5 over-predicts the total hydrogen mass generated for the FPT1 test. The hydrogen generation at the time of the peak oxidation phase for the FPT0 test is under-predicted while the hydrogen generation for the FPT1 test is over-predicted. In general, MAAP tends to over-predict mass relocations from the upper part of the bundle due to fuel rod collapses by the end of separate effects test transients.
CFD Analysis of Boron Distribution at Core Entrance in Case of Reverse Flow Induced From PRISE
Authors: Philipov Svetlin
Published: November 2014

Initiating events such as primary to secondary loss of coolant (PRISE) can lead to conditions forming reversed flow from the second to the primary circuit. Current issue shows the results of a CFD analysis of the distribution of boric acid on the entrance of the core in case of such reversed flow of coolant as a result of PRISE initiation event. Analyzed accident is included in the list of design basis accidents and requires precise approach in analyzing the phenomena associated with the possibility of injection of coolant with low concentration of boric acid in the primary side. The paper emphasizes on the application of CFD to solve the problem. Analyzing the accident is done in advance with the help of system code RELAP. The input data as flow rate, concentration and temperature at the inlet of the reactor is submitted as boundary conditions in FLUENT and boric acid mixing is analyzed to the core inlet.
Modeling and Simulation of Deaerator in Nuclear Power Plant
Authors: Xue Ruo-Jun , Sun Ji-Lin
Published: November 2014

Different deaerators have different structures, and the intensities of deaerating are different. But the deaerator models without considering the structure couldn’t show these aspects. In order to ensure the accuracy of simulation results, concrete structure of the deaerator should be taken into consideration. This paper carried out a mathematical model of horizontal type deaerator with constant nozzles and trays in nuclear power plants. It was built based on the structure of the segmentation deaerators and the process of heat transferring. This paper calculated the vapors condensation rate, it was based on heat transfer coefficient of direct contact condensation, and the heat transfer area of water film when working conditions were corresponding to the empirical formulas. While the working conditions were beyond the limits of empirical formulas, this paper would build models by conservation of energy, so the model could work under any working conditions. These models were installed in the simulation system of Qinshan Phase II., and were tested under variable power and accidental conditions. The testing results show that the models could fit for different steady working conditions. Compared with the outputs and the actual operating data, the error was small. Under the conditions of variable power and turbine tripping, changing of the parameters have the same trend curve with the actual operating data. Because of taking the concrete structure into consideration, these models can be performed better of the specific characteristics of the horizontal type deaerators with constant nozzles and trays in nuclear power plants, therefore the results are more accurate than the models without considering the structure.
Analysis of Effects of Buoyancy Force on Steam Condensation With Non-Condensable Gases in a Rectangular Channel
Authors: Wang Xian-Mao , Chang Hua-Jian , Han Li-Yong
Published: November 2014

With the development of science and technology, some important passive features have been used in nuclear reactors, one of which is passive containment cooling system (PCCS). In the system, steam condensation plays an important role in removing heat from the containment atmosphere during a postulated accident. It has been found that during most time of an accident, the gas regime in the containment will be under natural and mixed convection. Advanced pressurized water reactor (CAP1400), designed by State Nuclear Power Technology Corporation (SNPTC) in China, is one of Chinese national science and technology projects. Since the PCCS has been applied in CAP1400, the study of condensation with non-condensable gases under natural and mixed convection becomes necessary.

To have a deeper understanding on the phenomenon of condensation with non-condensable gases under natural and mixed convection, an experiment facility was set up by State Nuclear Power Technology Research & Development Centre (SNPTRD). The test section of the facility is a rectangular channel with one of the walls acting as a condensing plate. The effects of buoyancy force on steam condensation with non-condensable gases are investigated. Also, a CFD model is set up to simulate the process.
Prediction Model for Oxidation Layer Thickness in Flow Accelerated Corrosion
Authors: Han Ruixuan , Peng Yi , Li Huailin
Published: November 2014

A new prediction model for the oxidation layer thickness of carbon steel is developed, that is based on the parabolic time law of corrosion and the mass transport balance theory. The relationship between the oxidation layer thickness and temperature, pH, and flow velocity is discussed. The predicted results show that the oxidation layer thickness increases exponentially with increasing temperature and decreases exponentially with increasing flow velocity. The oxidation layer thickness increases with increasing pH until pH=10.5 and then decreases. The predicted results agree with experimental results.
Research on Two-Phase Flow Instability in Parallel Narrow Rectangular Channels Under Heaving Condition
Authors: Qian Libo , Gao Yingxian , Ding Shuhua
Published: November 2014

A theoretical model for flow instability in parallel rectangular channels under ocean conditions was established using lumped method based on the homogeneous model. The model consists of flow instability under static condition, the model of additional forces generated by ocean conditions, the constitutive and numerical models. The effect of ocean conditions was introduced through additional forces in the momentum equation generated by ocean conditions. The code was validated with the experimental data of flow instability under static conditions. Then the margin of flow instability, the threshold power and the periods of the system under static condition was calculated as a basis condition for comparison. The effect of the heaving condition on flow instability will be analyzed analytically. Besides these, the effect of asymmetric heating and throttling under heaving condition will be shown.
Elaboration of the System Based Code Concept — Activities in JSME and ASME: (3) Guidelines on Structural Reliability Evaluation for FBR
Authors: Takaya Shigeru , Machida Hideo , Kamishima Yoshio
Published: November 2014

This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the task force for the system based code in the Japan society of mechanical engineers in order to contribute to reducing differences in evaluated structural reliability by evaluators. They consist of five chapters, which are “General rules”, “Reliability evaluation”, “Failure scenario setting”, “Modeling”, and “Failure probability calculation”, respectively. In the chapter of “Reliability evaluation”, the general procedures of reliability evaluation are explained. Detailed procedures at each step are explained in the following chapters in the guidelines. Evaluation procedures for accumulation damage and crack propagation due to creep-fatigue interaction which is a typical degradation phenomenon in FBR are also provided in the appendix of the guidelines. In addition, some examples of random variables are prepared as standard input data for structural reliability evaluation.
Thermal-Hydraulic and Neutronic Analysis of a Re-Entrant Fuel Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors
Authors: Peiman W. , Pioro I. , Gabriel K.
Published: November 2014

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts one of which is the SuperCritical Water-cooled Reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as the coolant. The SCWR concept is considered to be an evolution of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), which comprise 81% of the current fleet of operating nuclear reactors and are categorized under Generation II nuclear reactors. The latter water-cooled reactors have thermal efficiencies in the range of 30–35% while the evolutionary SCWR will have a thermal efficiency of about 40–45%.

In terms of a pressure boundary SCWRs are classified into two categories, namely, Pressure Vessel (PV) SCWRs and Pressure Channel (PCh) SCWRs. A generic pressure channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350 and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve the thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel channel design (for PCh SCWR). Second, a nuclear fuel and fuel cycle should be selected. Third, materials for core components and other key components should be selected based on material testing and experimental results.

Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and re-entrant channel concepts. The objective of this paper is to study thermal-hydraulic and Neutronic aspects of a re-entrant fuel channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB which calculates the fuel centerline temperature, sheath temperature, coolant temperature and heat transfer coefficient profiles.

A lattice code and a diffusion code were used in order to determine the power distribution inside the core. Then, the heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents the fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature and the sheath temperature exceed the temperature limits of 1850°C and 850°C for fuel and sheath, respectively.
Prioritization of Underlying Precursors in Nuclear Accidents
Authors: Park Sujin , Yang Huichang , Heo Gyunyoung , Zubair Muhammad
Published: November 2014

The facts that the implicit precursors which are not easily quantified are underlying factors are already known. The current Probabilistic Safety Assessment (PSA) is limited in its ability to quantify the importance of accident causes. It is, therefore, difficult to achieve quantifiable decision-making for resource allocation. In this study, the methodology which facilitates quantifying these precursors and a case study is presented. First, four implicit precursors have been obtained by evaluating the causality and hierarchy structure of various accident factors. Eventually it turned out they represent the lack of knowledge. After four precursors are selected, sub-precursors were investigated and their cause-consequence relationship was implemented by Bayesian Belief Network (BBN). To prioritize the precursors, the prior probability is initially estimated by expert judgment and updated upon observations. The pair-wise importance between precursors is calculated by Analytic Hierarchy Process (AHP) and the results are converted into node probability tables of the BBN model. Using this method, the sensitivity and the posterior probability of each precursor can be analyzed so that it enables to make prioritization for the factors. Authors tried to prioritize the lessons-learned from Fukushima accident to demonstrate the feasibility of the proposed methodology.
Safety Requirements for Engineering Scale Pyroprocess Demonstration Facility
Authors: Moon Seong-In , Chong Won-Myung , You Gil-Sung , Ku Jeong-Hoe , Kim Ho-Dong
Published: November 2014

Pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems in Korea. The Korea Atomic Energy Research Institute has been studying pyroprocess technology, and the conceptual design of an engineering-scale pyroprocess facility, called the Engineering-Scale Pyroprocss Facility (ESPF), has been performed on the basis of a 10tHM throughput per year. In this paper, the concept of the ESPF was introduced and its safety requirements for the protection of facility workers, collocated workers, the off-site public, and the environment were introduced. For the identification of safety SSCs (Structures, Systems, and Components) and/or ACs (Administrative Controls), the following activities were conducted: 1) identifying hazards associated with operations, 2) identifying potential events associated with these hazards, and 3) identifying the potential preventive and/or mitigative controls that reduce the risk associated with these accident events. This study will be used to perform a safety evaluation for accidents involving any of the hazards identified, and to establish safety design policies and advance a more definite safety design.
Three Dimensional Coupled Simulation of Plutonium, Oxygen and Heat Diffusion in Mixed Oxide (U, Pu)O2 Fast Reactor Fuel
Authors: Liu Rong , Zhou Wenzhong , Yang Gang
Published: November 2014

Plutonium and oxygen diffusion with the high temperature gradient is one of the important fuel performance concerns in fast reactor (U, Pu)O2 fuel during irradiation, and will affect nuclear fuel materials properties, power distribution and overall performance of the fuel pin. This paper focuses on the plutonium, oxygen and heat diffusion within (U, Pu)O2 fast reactor fuel pellets. In this study, the correlations from the literature are used for thermal conductivity, specific heat, density, plutonium diffusion and oxygen diffusion. Three dimensional burnup dependent oxygen diffusion, plutonium diffusion and heat diffusion models are fully-coupled in steady states and transients to account for the effects on each other. The models are implemented into COMSOL Multiphysics to perform this analysis.
Investigation of MHD RANS Modeling Base on DNS Database Under the Advanced Blanket Design Conditions Utilized Molten Salt
Authors: Osawa Naoki , Yamamoto Yoshinobu , Kunugi Tomoaki
Published: November 2014

In this study, validations of Reynolds Averaged Navier-Stokes Simulation (RANS) based on Kenjeres & Hanjalic MHD turbulence model (Int. J. Heat & Fluid Flow, 21, 2000) coupled with the low-Reynolds number k-epsilon model have been conducted with the usage of Direct Numerical Simulation (DNS) database.

DNS database of turbulent channel flow imposed wall-normal magnetic field on, are established in condition of bulk Reynolds number 40000, Hartmann number 24, and Prandtl number 5.

As the results, the Nagano & Shimada model (Trans. JSME series B. 59, 1993) coupled with Kenjeres & Hanjalic MHD turbulence model has the better availability compared with Myong & Kasagi model (Int. Fluid Eng, 109, 1990) in estimation of the heat transfer degradation in MHD turbulent heat transfer.
Flow-Induced Vibration Suppression of Jet Pump in Boiling Water Reactor by Slip Joint Extension
Authors: Watanabe Masanobu , Kinugasa Kunihiko , Adachi Hiroyuki , Mori Hajime
Published: November 2014

Jet pumps are key components to feed cooling water into reactor core in boiling water reactor. Inside condition of jet pumps is high flow condition. Therefore, jet pumps have risk of damages by flow-induced vibration, especially, the leakage-flow-induced vibration at the slip joint between the inlet mixer and the diffuser in extended power uprating condition with increasing core flow rate or particular operating condition such as single loop operation that increases differential pressure of the slip joint. To mitigate the risk of the leakage-flow-induced vibration, slip joint extension which can be installed on the top of diffuser was developed (See Figure 1). Self-excited vibration is treated as negative damping i.e. unstable state. It is well-known that the leakage flow through divergent gap flow passage causes the negative damping. However, the configuration of the gap flow passage of the slip joint with slip joint extension is complicated flow passage which consists of convergent, divergent and parallel flow passage region. To addition to this, the leakage flow direction in normal or power uprating condition is opposite to in abnormal operating condition such as single loop operating. Therefore, it is necessary to identify the optimum configuration of gap flow passage of the slip joint extension to suppress leakage-flow-induced vibration for various operating conditions. To achieve this goal, the gap flow passage of the slip joint extension was determined using transfer matrix method based on the leakage-flow-induced vibration theory. The effect and characteristic of vibration suppression for the slip joint extension was confirmed by fundamental tests that simulated the slip joint configuration.
Application of 3D Coupled Code SPARKLE-2 to PWR Anticipated Transient Without Scram
Authors: Asano Koji , Sakamoto Hikaru , Imura Satoshi , Ogawa Junto
Published: November 2014

An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) followed by failure of the automatic reactor trip function of the reactor protection system. The failure of the reactor to shut down during the certain AOOs can lead to increase in reactor coolant system (RCS) pressure and decrease in departure from nucleate boiling ratio (DNBR) margin for a pressurized water reactor (PWR). Japanese standard PWRs are equipped with ATWS mitigation system which consists of a diverse mitigation system which is independent from the reactor trip system. The ATWS mitigation system automatically initiates isolation of the main steam line flow and the auxiliary feed water system under condition indicative of an ATWS.

Mitsubishi Heavy Industries, Ltd. (MHI) applies 3D coupled code, SPARKLE-2 [1] [2], to the ATWS evaluation. SPARKLE-2 is a 3D coupled code developed by MHI and consist of the PWR system transient analysis code M-RELAP5, the 3D neutron kinetics code COSMO-K [3] and the 3D core thermal-hydraulics code MIDAC [4]. SPARKLE-2 implements the 3D characteristics such as local moderator feedback and change in 3D power distribution during transient.

Thanks to gain from the 3D calculation, the analysis results show that the plant transients are effectively mitigated by the ATWS mitigation system and the RCS pressure and the minimum DNBR meet the safety criteria. These results also show that operational margins are increased, which enables more flexible design of the reload core.
Scaling Application of a Hot Leg SBLOCA Scenario to a Nuclear Power Plant
Authors: Querol Andrea , Gallardo Sergio , Verdú Gumersindo
Published: November 2014

Several experimental facilities, such as the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA), have been built to reproduce some accidental scenarios because full-scale testing is usually impossible to perform.

One of the objectives of these Integral Test Facilities (ITFs) is to obtain measured data to be compared to simulations in order to test the capability of the thermalhydraulic codes to reproduce experimental conditions.

The applicability of these experimental results to a full-size power plant system depends on the scaling criteria adopted.

The present paper is focused on the simulation and the scaling of the Test 1-2 in the frame of the OECD/NEA ROSA Project to a Nuclear Power Plant (NPP). This test simulates a hot leg 1% Small Break Loss-Of-Coolant Accident (SBLOCA) in a Pressurized Water Reactor (PWR) under the actuation of High Pressure Injection (HPI) system and Accumulator Injection System (AIS).

A scaled-up NPP TRACE5 input has been developed from a LSTF TRACE5 model validated by authors in previous works. The scaled-up model has been developed conserving the power-to-volume scaling ratios of LSTF components, initial and boundary conditions. Lengths and diameters of hot legs have been scaled from LSTF model trying to conserve Froude number.

A comparison between both TRACE5 models (LSTF and scaled-up NPP) is performed (system pressures, discharged inventory and collapsed liquid levels). Special TRACE5 models such as Choked flow model and OFFTAKE model have been tested. A 3D VESSEL component has been tested in comparison to 1D TEE component to simulate the hot leg where the SBLOCA is located and varying the break orientation (downwards and upwards). Finally, a sensitivity analysis has been made to determine the effect of the break size in the SBLOCA range.
Coolant Mixing in the VVER-440 Fuel Assembly Head
Authors: Romsy Tomas , Zacha Pavel
Published: November 2014

The issue of the temperature measurement in a nuclear reactor is an important element to ensure safe operation of the nuclear power plant. To prevent damages and radioactive releases the fuel in the reactor must be continuously cooled. The coolant temperature field of the VVER-440 reactor is measured with thermocouples installed at the outlet part of fuel assemblies. Since the power output of the fuel pins is not equal, a non-uniform temperature field at the inlet of the fuel assembly head is formed. Next, the temperature field is subsequently mixed by passing through the assembly head, which contains some constructional elements helping to mixing coolant flow. This mixing is not perfect and due to the effect described above, the signal on the thermocouple can be affected. This phenomenon was introduced in 2009 Atomic Energy Research (AER) Symposium in Bulgaria. 7 international institutions participated with the main goal to explain the mixing character of the coolant and to compare results. For further study of this phenomenon the new detailed computational model of the upper part of the fuel assembly was created and subsequently on the ANSYS FLUENT CFD code verified. The main output of these simulations is study of the coolant temperature distribution on the thermocouple. Computational model, based on the source geometry given by AER symposium, was created in preprocessor GAMBIT 2.4.6. Model contains over 13 million hexahedral cells. Thermohydraulic simulations where performed in ANSYS FLUENT v14.5 and results were compared with data from AER Symposium. There were considered two cases with different pins power outputs. With compare to AER symposium results the achieved resultant temperature on the thermocouple position for both cases indicate comparable accuracy. Furthermore, some flow fluctuations in the assembly head area where found.
Characterisation of FeCrAlY Coatings for Applications in Lead Environment
Authors: Di Gabriele Fosca , Houdkova Sarka , Soukup Ondrej , Karnik Dalibor
Published: November 2014

The use of heavy liquid metals (HLM), such as Lead, Pb, is being considered as a coolant in Generation IV fast Reactors. However, structural materials suffer significant damage when in contact with HLMs. Both austenitic and ferritic-martensitic steels are considered and are susceptible to corrosion and/or degradation of mechanical properties. One of the approaches to mitigate erosion/corrosion problems is the use of coatings. This is a layer of a different chemical composition between the base material and the liquid metal coolant in order to prevent the bulk metal from corrosive effects. Coatings are proposed as a valid protection against high temperature damage in this environment. Their capability to grow more stable and protective oxides, by introducing the oxide forming elements (e.g. Al, Si,…) in higher amount, is proven to be an effective alternative to material engineering.

For this specific application, several coatings deposition techniques and compositions have been proposed and tested. In this work the High Velocity Oxygen Fuel, HVOF, combined with laser melting was selected for deposition of FeCrAlY coatings. The combination of the two technologies lead to a compact and adherent coating with an enriched content of Al. The method was evaluated in terms of the corrosion resistance of the coating and also its effect on the microstructure of the substrate alloy. Several attempts were carried out to modify spraying and laser parameters in order to minimise the effect on the substrate and keep the protective properties.

Results are discussed in terms of deposition parameters and protection characteristics.
Investigation of Critical Heat Flux in Single Fuel Pin With and Without Wire Spacer
Authors: Le Tri Dan , Inaba Noriaki , Takahashi Minoru
Published: November 2014

Light water reactor could have fast neutron spectrum with high conversion ratio nearly equal unity by using tight lattice fuel assembly with wire spacer. There were some previous about critical heat flux for tight lattice but it were not focused on small range of qualities and also not directly using water as a coolant. We experimentally simulated vertical single fuel rod geometry with and without a wire spacer by using an electrically heated stainless steel rod. The rod is cooled by single or two-phase water in vertical up flow (from the bottom to the top) depending on the electrical input. We determined the critical heat flux for this system by varying the inlet temperature from 333 to 373 K and mass fluxes from 205 to 410 kg/m2s. The result show the critical heat flux (CHF) data in two phase flow condition base on inlet and outlet condition in both case of heater pin with and without wire. The CHF values were higher with wire than without wire due to the effect of wire and spiral flow.
Sloshing Suppression Method for a Rectangular Pool Using Immersed Blocks and Baffle Plates
Authors: Watanabe Urara , Niwa Hiroshi , Tsuyuki Akira , Warashina Masahiko , Ito Ryo , Watanabe Masanobu
Published: November 2014

There is a possibility that water in a spent fuel storage pool may overflow due to sloshing during long-period earthquakes. Therefore, this paper presents two sloshing suppression methods for a rectangular pool to reduce the volume of overflow water. Vibration tests were carried out to evaluate the volume of overflow water. The 1/20-scale model pool is used. First method is applying immersed blocks on the bottom of the rectangular pool. The volume of the water over the sidewall should be the maximum when the 1st sloshing mode is excited, and this behavior has significant influence on the volume of overflow water. The immersed blocks suppress the 1st sloshing mode, thereby reducing the over flow water. Vibration test were conducted by changing the following conditions: height of blocks, open area ratio, block position in the excitation direction, and number of opening sections. Changing of the natural frequencies and the amplification ratios are confirmed by the sinusoidal sweep test. In random wave excitation tests, the volume of overflow water from the pool with the best configuration blocks is lower than about 60% of that from the regular pool. Second method is applying horizontal baffle plates on the sidewall of rectangular pool. Horizontal baffle plate is a well-known sloshing suppression method. However, there is a little information about the relationship between installation condition of baffle plate and volume of overflow water Vibration test were conducted by changing the following conditions: installation height level, overhang length of baffle plate. In random wave excitation tests, the volume of overflow water from the pool with the best installation condition baffle plate is lower than 40% of that from the regular pool.
A CFD Analysis for the Advanced Accumulator
Authors: Lu Donghua , Wang Kuo , Su Qianhua , Xing Jun
Published: November 2014

A new design for the flow control component of an advanced accumulator (ACC) was introduced and numerically investigated in the paper. The flow control component can produce high flow rate and low flow rate at the different stages during the safety injection. The FLUENT computational fluid dynamics (CFD) was used in the simulation of the flow pattern in the flow chamber and the outlet. The flow pattern and pressure gradient in the vortex chamber were investigated. The results show the design can realize the high flow and low flow in this design.
Experimental Studies on Critical Heat Flux in Tight Lattice Rod Bundles
Authors: Lang Xuemei , Gong Houjun , Zhou Lei , Xie Feng , Liu Ye
Published: November 2014

The tight fuel lattice of pressurized water reactors (PWR) is helped to reduce the volume ratio of water-uranium, to increase the conversion ratio, to decrease the volume of core. It is especially useful for very high burnup and high volume power flux. The design of tight-lattice pressurized water reactors requires the knowledge of critical heat flux (CHF) in tight rod bundles. The tight hexagonal 19-rod bundles is used in this test. There are 4 wires wrapped in outside wall of each rod to support and locate. Experimental investigations on CHF behavior in the two kind bundles of helix angle 3° and 5° were performed. The CHF data points have been obtained in a range of parameters: pressure 8.0–16.6 MPa, mass flux 164.6–3283.0 kg/m2s and bundle exit steam quality −0.315 to 0.747. It is found that the CHF value of helix angle 5° bundle was more higher than that of helix angle 3° bundle in the same T/H condition. The effect of different parameters on CHF in the tight rod bundle is similar to that in the open literature. The CHF correlations of helix angle 5° bundle was obtained based on the test data.
Delayed Equilibrium Model (DEM) of Flashing Choked Flows Relevant to LOCA and Implementation in System Codes
Authors: Bartosiewicz Yann , Seynhaeve Jean-Marie
Published: November 2014

In the context of nuclear reactor safety, a pipe breach in the primary circuit is the initiator of a Loss of Coolant Accident (LOCA). The calculation of leak rates involving the discharge of water and steam mixtures plays an important role in the modeling of LOCA’s for both GEN II and GEN III reactors, and also for the Supercritical Water Reactor of GEN IV. Indeed, the flow though the breach determines the depressurisation rate of the system and the time to core uncover which in turn are of major concern for when and how different mitigation auxiliary systems will be initiated and be efficient.

This paper deals with the new development of the DEM model focused on thermodynamic non-equilibrium conditions, which prevail in the flashing flow process near the critical section. This model, developed at the University of Louvain (UCL), is the 1-D Delayed Equilibrium Model (DEM) for choked or critical flow rate in steady state or quasi-steady state conditions. The DEM are assessed against experimental data such as Super Moby-Dick and BETHSY experiments done in CEA during the eighties. The DEM model has been recently implemented in the WAHA code, which is based on a two fluid 1D six equations model. The methodology can be applied to other system code (CATHARE, RELAP, etc.).
Implementation of an Approach for Characterisation of Radiological Inventory for Decommissioning at Chinon A2 NPP
Authors: Traichel A. , Tardy F. , Mummert M.
Published: November 2014

A general overview of the existing radioactive inventory in the plant is necessary for the decommissioning of nuclear power plants. Based on the knowledge about radiological inventory, appropriate decommissioning techniques and procedures can be specifically used. In order to derive the existing radiological activity in the facility a study was carried out to obtain a representative overview of the total radiological situation at the NPP. Within a study a generic methodology for the radiological characterization was developed. This methodology has been applied on the CO2-circuit of the gas-cooled, graphite-moderated reactor Chinon A2 (MAGNOX type). This paper covers the implementation of an approach for characterisation of radiological inventory for decommissioning.

The approach aims at the definition of the number and distribution of local sampling, required measurements as well as suitable measurement systems leading to a confident result with minimized effort in sampling. The paper covers two main objectives: 1. Methodology at and 2. Determination of radiological inventory based on measured data. The proposed methodology is a stepwise procedure which offers the possibility for minimizing the number of required measurements/sample analyses. At the first step the underlying system is an “as-simple-as-possible”-example with homogeneous contamination. In a second step the methodology is expanded to a more realistic and complex system, for which additional investigations have to be performed. The determination of the radiological inventory using the methodology has to consider a given confidence level and maximum allowed error. Therefore statistical assessment is widely used in estimations.

The result of this first part of study generates the basis for further investigation. This comprises application of methodology to the mentioned technical system. Therefore corresponding measurement and analysis data have to be delivered and proven regarding adequacy for the proposed methodology. From the dataset various measurement systematic and retained parameters could be derived. The accuracy of given measured data was checked by further examination. The result of the performed analysis leads to a statement about the activity in the primary circuit. The result of this study is an comprehensive estimation of the activity by defined statistical processing of analysed data. The result consists moreover of the analysis of the measurement plan and of distribution and deviation within the measured data. Suggestions for further measurement campaigns are provided based on the deviations and inconsistencies of the data. With the help of these suggestions it should be possible to decrease the number of samples and measuring data as well as improve the comparability of separate measurement processes. Particular potential for improvement of the result for inventory can be seen in a deeper analysis of uncertainties, this was realised and will be explained in the paper.
Spent Fuel Canister Criticality Safety Calculation in Groundwater Immersion Accident
Authors: Wu Siyuan , Yang Bo , Wu Hexi , Wei Qianglin , Liu Yibao
Published: November 2014

This paper studies the KBS-3 spent fuel canister criticality safety issue in the event of a groundwater penetration accident. First, the study calculates the composition of the PWR spent fuel assemblies by using MCBURN software, and the coupling of the MCNP and ORIGEN 2.1 computer codes. Then, with the help of Isotope Generation and Depletion Code, it calculates the composition of various types of actinides and their daughters for 100,000 years. Finally, the above calculation results are used in MCNP5 to calculate the effective multiplication factor keff of the canister for different degrees of groundwater penetration. The study finds that the maximum keff of the canister is 0.79609 in groundwater penetration accident, satisfying criticality safety standard.
Modelling of RBMK Spent Nuclear Fuel Characteristics
Authors: Smaizys Arturas , Poskas Povilas , Narkunas Ernestas
Published: November 2014

After the final shutdown of Ignalina NPP, total amount of spent nuclear fuel is approximately 22 thousands of fuel assemblies. Radionuclide content and its characteristics in spent fuel are initial data for analysis of various safety related areas such as shielding, thermal analysis, radioactive releases and other processes. Experimental investigations of radionuclide content and characteristics in spent nuclear fuel are complicated and expensive, therefore numerical evaluation methods are widely used. Numerical modelling of spent RBMK fuel characteristics was performed using TRITON code from SCALE 6.1 system. Activities of fission products and actinides, gamma and neutron sources, decay heat obtained with TRITON code are compared with previous modelling results obtained using SAS2H sequence from the former SCALE 4.3 version. Some evaluated parameters are compared with published experimental data for RBMK spent nuclear fuel.
Fast Neutron-Flux Monitoring Instrumentation for Lead Fast Reactors: A Preliminary Study on Fission Chamber Performances
Authors: Lepore Luigi , Remetti Romolo , Cappelli Mauro
Published: November 2014

Although Sodium Fast Reactors (SFRs) are the most investigated solutions for the future fast-flux facilities so far, Lead Fast Reactors (LFRs) promise to be a very competitive alternative thanks to their peculiarity concerning coolant-safety, fuel cycle and waste management.

Nevertheless, the development of LFRs presents today some drawbacks still to be solved. Due to the harder neutron flux, the current instrumentation developed for SFRs is likely to be extended to LFRs as a first attempt. Otherwise, new monitoring instrumentation could be developed in order to assure more tailored results. Different measurement technologies can be considered for fast flux monitoring and flux absolute measurements in order to provide a reliable and quick calibration of the overall reactor neutron instrumentation. The goal of this paper is to study the validity of typical fast reactor fission chamber designs (e.g. SuperPhénix fission chambers), indicating which are the limitations when used in a LFR environment. Afterwards, alternative detector solutions with enhanced sensitivity and response will be proposed.
Testing of Ceramics for Production of Heating Elements for Usage in High Temperature Helium Loop
Authors: Berka Jan , Rotek Antonín , Vit Jan , Kutzendorfer Jaroslav
Published: November 2014

Heating elements are one of the important components of High Temperature Helium Loop. By means of heating elements the heating of test section of the loop to required temperature is provided even in high gas flow rates in the loop. The insulation of heating elements was manufactured of cordierite ceramics. With these elements the maximum projected temperature 900 °C could not be reached because of decreasing of electric resistance among the spirals. Therefore several tests were done in order to discover the reason of the problem. Consequently properties of new ceramic material based on corundum for manufacturing of the new heating elements were verified.
Experimental Measurements and CFD Evaluation of the Onset Flow Boiling at Low Pressure and High Subcooling Flow of R-11 Within Vertical Annulus
Authors: Bouaichaoui Y. , Kibboua R. , Matkovič M.
Published: November 2014

The knowledge of the onset of subcooled boiling in forced convective flow at high liquid velocity and subcooling is of importance in thermal hydraulic studies. Measurements were performed under various conditions of mass flux, heat flux, and inlet subcooling, which enabled to study the influence of different boundary conditions on the development of local flow parameters. Also, some measurements have been compared to the predictions by the three-dimensional two-fluid model of subcooled boiling flow carried out with the computer code ANSYS-CFX-13.

A computational method based on theoretical studies of steady state two phase forced convection along a test section loop was released. The calculation model covers a wide range of two phase flow conditions. It predicts the heat transfer rates and transitions points such as the Onset of Critical Heat Flux.