Thermal–hydraulic and thermo-structural analysis of first wall for Indian DEMO blanket module

The first wall (FW) is one of the most important components of any fusion blanket design. India has developed two concepts of breeding blanket for the DEMO reactor: the first one is Lead–Lithium cooled Ceramic Breeder (LLCB), and the second one is Helium-Cooled Ceramic Breeder (HCCB) concept. Both the concept has the same kind of FW structure. Reduced Activation Ferritic Martensitic steel (RAFMS) used as the structural material and helium (He) gas is used to actively cool the FW structure. Beryllium (Be) layer of 2 mm is coated on the plasma side of the FW as the plasma facing material. Cooling channels running in radial–toroidal–radial direction in the RAFMS structure are designed to withstand the maximum He pressure of 8 MPa. Heat transfer coefficients (HTC) obtained form the correlations revealed that required cooling could be achieved by artificially roughened surface towards the plasma-side wall of He cooling channel which helps to keep the RAFMS temperatures below the allowable limit. A 1D analytical and 2D thermal–hydraulic simulation studies using ANSYS has been performed based on the heat load obtained from neutronics calculations to confirm the heat removal and structural integrity under various conditions including ITER transient events. The required helium flow through the cooling channels are evaluated and used to optimize the suitable header design. The detail design of FW thermal–hydraulics, thermo-structural analyses, and He flow distribution network will be presented in this paper.     

Thermal–hydraulic and safety analysis for Chinese helium-cooled solid breeder TBM cooling system

The thermal–hydraulic behavior and safety performance of the Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system (HCS) has been studied using RELAP5/Mod3.4 code. According to accident analysis specification for TBM, two design basis accidents including loss of off-site power and TBM first wall (FW) ex-vessel coolant pipe break are investigated. The influences of different break locations and plasma termination behaviors are analyzed comprehensively. The results show that natural circulation is established in helium cooling circuit and the TBM can be cooled effectively after loss of off-site power. It is much more critical when the pipe break occurs at the downstream side of the circulator compared with that of upstream side of the circulator. In case of a more serious accident that the ex-vessel break extends to the TBM FW, the results reveal that TBM could be cooled down by natural circulation and radiation. In addition, at the beginning of ex-vessel loss of coolant accident (LOCA), large temperature difference between break and intact TBM FW pipes is found. The accidental results finally show that the integrity of the FW can be guaranteed if the plasma is terminated with a 3 s delay time by fusion power shutdown system (FPSS) in the case of ex-vessel LOCA.     

Implementation of high-fidelity neutronics and thermal–hydraulic coupling calculations in HNET

To perform nuclear reactor simulations in a more realistic manner, the coupling scheme between neutronics and thermal-hydraulics was implemented in the HNET program for both steady-state and transient conditions. For simplicity, efficiency, and robustness, the matrixfree Newton/Krylov (MFNK) method was applied to the steady-state coupling calculation. In addition, the optimal perturbation size was adopted to further improve the convergence behavior of the MFNK. For the transient coupling simulat...     

Development of a thermal–hydraulic analysis code for research reactors with plate fuels

The thermal–hydraulic analysis code THAC-PRR has been developed with Visual Fortran 6.5 for the investigation of plate type fuel reactors. It is based on the fundamental conservation of mass, momentum and energy, and proper constitutive correlations for flow friction factor, heat transfer and thermophysical properties. Moreover, a simple and improved lumped-differential method has been adopted to analyze the conjugate heat transfer between the fuel plate and the coolant. The Reactivity Insertion Accident (RIA) and Loss Of Flow Accident (LOFA), which have been defined in the IAEA 10 MW MTR Benchmark program, were analyzed with this developed program for the code-to-code validation. Good agreement was achieved. Furthermore, the accidents due to the partial (95%) and total (100%) blockage of one channel in the IAEA 10 MW MTR were investigated with THAC-PRR. The results showed that if the blockage occurred in the average channel, there was no boiling occurred even the channel was totally obstructed. The reason was that the heat was transferred to the adjacent channels by conduction through the fuel plates which formed the obstructed channel. However, if the blockage occurred in the hot channel, boiling did occur. This indicated that it is very important to consider the interaction between the blocked channel and the adjacent channels in this type of transient.     

Quantitative functional failure analysis of a thermal–hydraulic passive system by means of bootstrapped Artificial Neural Networks

The estimation of the functional failure probability of a thermal–hydraulic (T–H) passive system can be done by Monte Carlo (MC) sampling of the epistemic uncertainties affecting the system model and the numerical values of its parameters, followed by the computation of the system response by a mechanistic T–H code, for each sample. The computational effort associated to this approach can be prohibitive because a large number of lengthy T–H code simulations must be performed (one for each sample) for accurate quantification of the functional failure probability and the related statistics.     

Coupled neutronics/thermal–hydraulics analysis of CANDU–SCWR fuel channel

At supercritical pressure condition, the thermal–hydraulics behavior of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal–physical properties across the pseudo-critical line. A coupling analysis of neutronics and thermal–hydraulics has become important for SCWR, because of the strong link between the water density and the neutron spectrum and subsequently the power distribution. The neutronics code Monte Carlo N-Particle code (MCNP) and the subchannel code Advanced Thermal–Hydraulics Analysis Subchannel (ATHAS) are used in a coupled way to better understand the design characteristics of a pressure tube type SCWR fuel channel. The results show that: the developed coupled code system can be used to analyze pressure tube type SCWR fuel bundles; improved radial fuel enrichment profile will optimize the coolant and cladding temperature distribution to meet the design criteria; smaller pressure tube pitch will result in more flatten axial power distribution and more uniform radial power distribution.     

Dynamic scenario quantification for level 2 PRA of sodium-cooled fast reactor based on continuous Markov chain and Monte Carlo method coupled with meta-model of thermal–hydraulic analysis

In probabilistic risk assessment (PRA), an event tree (ET) methodology is widely used to quantify accident scenarios which result in core damage and fission products release. However, the current approach using the ET methodology is not applicable to evaluate dynamic characteristics of accident progression, when the accident progression is time-dependent and headings in the ET have inter-dependency between events. Thus, a dynamic approach of accident scenario quantification is necessary to evaluate more realistic PRA.

This research addressed this need by developing a dynamic scenario quantification method for the level 2 PRA by coupling of Continuous Markov chain and Monte Carlo (CMMC) method and a plant thermal–hydraulic analysis code for a sodium-cooled fast reactor (SFR).

The CMMC method is applied to protected loss of heat sink (PLOHS) accident of the SFR to analyze dynamic scenario quantifications. The coupling method requires heavy computational cost and it makes difficult to quantify the whole accident scenarios by comparing the results from existing plant state analysis codes. Thus, a meta-analysis coupling method is proposed to obtain dynamic scenario quantifications with reasonable computational cost. Also, a categorizing method is used to depict analytical results in a transparent manner.     

Neutronics analysis of the power flattening and minor actinides burning in a thorium-based fusion–fission hybrid reactor blanket

A neutronics analysis has been performed for a thorium fusion breeder with a special task of burning minor actinide ²³⁷Np, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm, and production of ²³³U for the future PWR application. Under a first wall fusion neutron wall loading of 0.1 MW/m² by a plant factor of 100%, preliminary neutronics calculations have been performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. To obtain a quasi-constant nuclear heat production density, 11 fuel rods containing the mixture of ThO₂ and minor actinides are placed in a radial direction in the fissile zone where ThO₂ is mixed with variable amounts of minor actinides. Calculation results show that the tritium breeding ratio is greater than 1.05 for both investigated Cases A and B, and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those models during the operation period. The blanket energy multiplication factor M, varies between 13.8 and 29.6 depending on the fuel types at the end of the operation period. The peak-to-average fission power density ratio (Γ) is less than 1.66 and 1.68 for both Cases A and B, respectively during the operation time. After 720 days of plant operation, the accumulated ²³³U is 1277 and 1725 kg in the blanket for the Cases A and B, respectively.     

CFD investigation of flow inversion in typical MTR research reactor undergoing thermal–hydraulic transients

Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.     

Thermal–hydraulic analysis of space nuclear reactor TOPAZ-Ⅱ with modified RELAP5

With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.     

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO₂ fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.     

Nuclear and thermal–hydraulic characteristics for an LMR core fueled with 20% enriched uranium metallic fuel

As a part of the core design development of KALIMER (150 MWe), the KALIMER core was initially designed with 20% enriched uranium metallic fuel. In this core design, the primary emphasis was given to realize the metallic fueled core design to meet the specific design requirements; 20% and below uranium enrichment and a minimum fuel cycle length of one year. The core was defined by a radially homogeneous core configuration incorporated with several passive design features to give inherent passive means of negative reactivity insertion. The core nuclear performance based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and maximum discharge burnup of 47.3 MWD/kg. When comparing with conventional plutonium metallic fueled cores of the same power level, the present uranium metallic fueled core has a lower power density due to its increased physical core size. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. The transition from the uranium startup to equilibrium cycle is feasible without any design change. Core nuclear performance characteristics in the present core design are attributed to the specific design requirements of enrichment restriction and fuel cycle length.     

Three-dimensional FE analysis of the thermal–mechanical behaviors in the nuclear fuel rods

In order to implement numerical simulation of the thermal–mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal–mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal–mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm³, the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm² K to 0.01 W/mm² K, while increases up to 54.7% when h decreases from 0.01 W/mm² K to 0.005 W/mm² K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal–mechanical behaviors in the fuel rod; when the gap size varies from 0.03 mm to 0.1 mm, the highest temperature in the pellet increases by 19.7%, and the maximum first principal stress at the outer pellet surface decreases by 17.4%; it is critical to optimize the gap size in order to reduce the pellet–cladding mechanical interaction and avoid their contact at early stage. This study lays a foundation for the further research on the irradiation-induced mechanical behaviors in the fuel rods.     

Electromagnetic–thermal–structural coupling analysis of the ITER edge localized mode coil with flexible supports

In a fusion reactor, the edge localized mode(ELM) coil has a mitigating effect on the ELMs of the plasma. The coil is placed close to the plasma between the vacuum vessel and the blanket to reduce its design power and improve its mitigating ability. The coil works in a high-temperature,high-nuclear-heat and high-magnetic-field environment. Due to the existence of outer superconducting coils, the coil is subjected to an alternating electromagnetic force induced by its own alternating current and the outer magnetic field. The design goal for the ELM coil is to maintain its structural integrity in the multi-physical field. Taking as an example the middle ELM coil(with flexible supports) of ITER(the International Thermonuclear Fusion Reactor), an electromagnetic–thermal–structural coupling analysis is carried out using ANSYS. The results show that the flexible supports help the three-layer casing meet the static and fatigue design requirements. The structural design of the middle ELM coil is reasonable and feasible. The work described in this paper provides the theoretical basis and method for ELM coil design.     

On monitoring the tritium breeder in a lead–lithium cooled ceramic breeder (LLCB) module of the ITER

The operation of a tritium breeder is a most process among engineering problems of DEMO. In this study, a design for monitoring tritium-breeding in the reactor is discussed. Additionally, a system for the experimental estimation of the tritium-breeding ratio (TBR) and the tritium-breeding dynamics in a lead–lithium cooled ceramic breeder (LLCB) test module used in the ITER is proposed. The systems are based on tritium and neutron-flux measurements under the ITER plasma D–T experiments and the use of lithium ortho-silicate and lithium carbonate samples and neutron detectors. Different lithum-6 and lithium-7 isotope contents in the samples are used to measure neutron spectrum. The samples and detectors are delivered in containers to the test breeder module (TBM) on a monitor channel connecting the TBM to an operating zone of the ITER. The tritium content in the samples is measured in a laboratory by the liquid scintillation method.Pneumatic control is used to deliver the samples to the TBM and to extract the samples using the channel during plasma-operational pauses. Neutron calculation is performed to estimate the tritium content in the samples and the heat distribution in the materials of the channel under reactor irradiation. A measurement accuracy of the tritium content in the carbonate and orthosilicate samples can attain a level of 7% and 10%, respectively. The results of the channel-cooling calculation performed under the nominal operating conditions of the TBM (a plasma pulse) are presented in the paper.     

An APXPS endstation for gas–solid and liquid–solid interface studies at SSRF

In the past few decades, various surface analysis techniques find wide applications in studies of interfacial phenomena ranging from fundamental surface science,catalysis, environmental science and energy materials.With the help of bright synchrotron sources, many of these techniques have been further advanced into novel in-situ/operando tools at synchrotron user facilities, providing molecular level understanding of chemical/electrochemical processes in-situ at gas–solid and liquid–solid interfaces.Designing a proper endstation for a dedicated beamline is one of the challenges in utilizing these techniques efficiently for a variety of user's requests. Many factors,including pressure differential, geometry and energy of the photon source, sample and analyzer, need to be optimized for the system of interest. In this paper, we discuss the design and performance of a new endstation at beamline02 B at the Shanghai Synchrotron Radiation Facility for ambient pressure X-ray photoelectron spectroscopy studies.This system, equipped with the newly developed hightransmission HiPP-3 analyzer, is demonstrated to be capable of efficiently collecting photoelectrons up to 1500 eV from ultrahigh vacuum to ambient pressure of 20 mbar.The spectromicroscopy mode of HiPP-3 analyzer also enables detection of photoelectron spatial distribution with resolution of 2.8 ± 0.3 lm in one dimension. In addition,the designing strategies of systems that allow investigations in phenomena at gas–solid interface and liquid–solid interface will be highlighted through our discussion.     

Human reliability analysis in the man–machine interface design review

Advanced, computer-based man-machine interface (MMI) is emerging as part of the new design of nuclear power plants. The impact of advanced MMI on the operator performance, and as a result, on plant safety should be thoroughly evaluated before such technology is actually adopted in the plants. This paper discusses the applicability of human reliability analysis (HRA) to support the design review process. Both the first-generation and the second-generation HRA methods are considered focusing on a couple of promising HRA methods, i.e. ATHEANA and CREAM, with the potential to assist the design review process.     

GISAXS and GIWAXS analysis of amorphous–nanocrystalline silicon thin films

Amorphous–nanocrystalline silicon thin films were deposited by Plasma Enhanced Chemical Vapor Deposition (PECVD) on glass substrate with various silicon nano-crystal size distributions and volume fractions. The samples were examined by Grazing Incidence Small Angle X-ray Scattering (GISAXS) and Grazing Incidence Wide Angle X-ray Scattering (GIWAXS) at the Austrian SAXS beamline (Synchrotron Elettra, Trieste) using an X-ray beam energy of 8 keV. The grazing incidence angle varied from the critical angle to 0.2° above the critical angle. This allowed the examination of the samples at different depths, and the distinction of the surface scattering contribution from the particles scattering in the bulk. The sizes of the “particles” obtained from the horizontal and vertical sections of 2D GISAXS patterns were between 2 nm and 6 nm. Since GISAXS is sensitive to electron density differences (contrast) between the scattering bodies and the surrounding matrix, it is not evident whether the particles are nano-crystals or just voids embedded in amorphous matrix. However, the size of the crystals calculated from the line-shape analysis of peaks in GIWAXS spectra and the crystal size distribution obtained from High-Resolution Transmission Electron Microscopy (HRTEM) images agree well with the size of “particles” estimated from GISAXS, strongly indicating that the observed particles are silicon nano-crystals.     

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO₂–ThO₂ fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO₂–ThO₂ fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO₂–UO₂ fuel pins are employed to achieve long-cycle length of ～4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods.