Overview of design and thermal–hydraulic analysis of Indian solid breeder blanket concept

India has developed two concepts of breeding blanket for the DEMO reactor: one is Lead Lithium Ceramic Breeder (LLCB), and the other one is Helium-cooled Ceramic Breeder (HCCB) concept. Indian HCCB concept is having edge on configuration of helium-cooled solid breeder with RAFMS structure. Li₂TiO₃/Li₄SiO₄ and beryllium are used as the tritium breeder and neutron multiplier, respectively. 2D thermal–hydraulic simulation studies using ANSYS have been performed based on the heat load obtained from neutronics calculations to confirm heat removal under ITER pulsed operation. Transient thermal analysis has been simulated in ANSYS for the ITER relevant operational conditions. Thermal analysis provides important information about the temperature distribution in different materials used and their temperature–time histories. Result of thermal–hydraulic simulations shows that in each cycle, the maximum temperature of all materials remains same. The peak temperatures of all materials are well within their limiting value. Concept designs of HCCB blanket and its thermal hydraulic analysis 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.     

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.     

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.     

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.     

Analysis of high β regimes for DEMO

Stability limit calculations are presented for a range of tokamak power plant equilibria. The current drive requirements to sustain the optimised equilibrium profiles are confirmed by a transport code and the plasma shape is obtained from free-boundary equilibrium calculations. A pressure pedestal is included according to empirical scaling and ballooning mode stability limits. A terative optimisation of the profiles is undertaken to improve the baseline profiles in order to achieve the highest possible plasma performance and most favourable magnetohydrodynamic stability within conservative assumptions in order to increase confidence in the availability and control of the plasma. This results in a fully noninductive baseline operating scenario for a tokamak power plant design which has a broad low-shear q-profile which is meant to complement previous advanced tokamak design studies.     

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.     

Thermal—hydraulic Optimization of Water—cooled Center COnductor Post for Spherical Tokamaks Reactor

This paper proposes a conceptual structure of segmental water-cooled Center Conductor Post(CCP) to be flexble in installment and replacement.Thermal-hydraulic optimization and sensitivity analysis of key parameters are performed based on a reference fusion transmutation system with 100MW fusion power.Numerical simulation by using a commercial code PHOENICS has been carried out to be close to the thermal-hydraulic analytical results of the CCP mid-part.     

Thermal and mechanical properties of hydrides of Zr–Hf alloys

Polycrystalline bulk samples of δ-phase Hf hydrides with various Zr contents were prepared and their high-temperature stability and thermal and mechanical properties were investigated. The phase structure was examined between room temperature and 973 K using high-temperature X-ray diffraction and thermogravimetric–differential thermal analysis. From room temperature to 673 K, the coefficient of linear thermal expansion, specific heat capacity, and thermal conductivity were evaluated. The Vickers hardness and sound velocity were measured at room temperature, and the elastic modulus was evaluated. The effect of the Zr content on the high-temperature stability and the thermal and mechanical properties of Hf hydrides was studied.     

NAPTH: Neutronics analysis code system for the fusion–fission hybrid reactor with pressure tube type blanket

The fusion–fission hybrid reactor is considered as a potential path to the early application of fusion energy. A new concept with pressure tube type blanket has recently been proposed for a feasible hybrid reactor. In this paper, a code system for the neutronics analysis of the pressure tube type hybrid reactor is developed based on the two-step calculation scheme: the few-group homogeneous constant calculation and the full blanket calculation. The few-group homogeneous constants are calculated using the lattice code DRAGON4. The blanket transport calculation is performed by the multigroup Monte Carlo code. A link procedure for fitting the cross sections is developed between these two steps. An additional procedure is developed to calculate the burnup, power distribution, energy multiplication factor, tritium breeding ratio and neutron multiplication factor. From some numerical results, it is found that the code system NAPTH is reliable and exhibits good calculation efficiency. It can be used for the conceptual design of the pressure tube type hybrid reactor with precise geometry.     

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...     

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.     

Thermal—hydraulic stability of a natural circulation system with nuclear feedback

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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.     

Fast pulse sampling module for real-time neutron–gamma discrimination

An adaptable and compact fast pulse sampling module was developed for the neutron–gamma discrimination. The developed module is well suited for low-cost and low-power consumption applications. It is based on the Domino Ring Sampler 4(DRS4) chip, which offers fast sampling speeds up to 5.12 giga samples per second(GSPS) to digitize pulses from front-end detectors. The high-resolution GSPS data is useful for obtaining precise real-time neutron–gamma discrimination results directly in this module. In this study, we have implemented real-time data analysis in a field programmable gate array. Real-time data analysis involves two aspects: digital waveform integral and digital pulse shape discrimination(PSD). It can significantly reduce the system dead time and data rate processed offline. Plastic scintillators(EJ-299-33), which have proven capable of PSD, were adopted as neutron detectors in the experiments. A photomultiplier tube(PMT)(model #XP2020) was coupled to one end of a detector to collect the output light from it. The pulse output from the anode of the PMT was directly passed onto the fast sampling module. The fast pulse sampling module was operated at 1 GSPS and 2 GSPS in these experiments, and the AmBe-241 source was used to examine the neutron–gamma discrimination quality. The PSD results with different sampling rates and energy thresholds were evaluated. The figure of merit(FOM) was used to describe the neutron–gamma discrimination quality. The best FOM value of 0.91 was obtained at 2 GSPS and 1 GSPS sampling rates with an energy threshold of 1.5 MeV_(ee)(electron equivalent).     

Thermal conductivity and electrical resistivity of liquid Ag–In alloy

To understand and predict the progression of core meltdown accidents in nuclear power plants, it is important to understand the behavior of molten core materials. We focused on the melting behavior of Ag–In–Cd alloys used in the control rods of pressurized water reactors that are known to melt first when a severe accident occurs. To obtain fundamental knowledge about these alloys, we studied the thermal conductivity of Ag–In binary alloys in this study. We evaluated thermal conductivity using two approaches: evaluating from thermal diffusivity up to 873 K measured by the laser-flash method, and calculating based on the Wiedemann–Franz law using the electrical resistivity up to 1273 K measured by the four-probe method. The values of thermal conductivity of liquid Ag–In alloys obtained by these two methods agreed well except for pure indium. Although Ag is known as a material that has one of the highest thermal conductivities, the thermal conductivity of liquid Ag–In alloys is much lower than that of pure liquid Ag (177 W/mK at 1273 K), but almost the same or less than that of liquid In (59.2 W/mK at 1273 K) in all Ag_1?xIn_x (x = 0.2–0.8) alloys at all temperatures in this measurement.     

Thermal–hydraulic modeling of water/Al2O3 nanofluid as the coolant in annular fuels for a typical VVER-1000 core

In this paper, a thermal–hydraulic analysis of nanofluid as the coolant is performed in a typical VVER-1000 reactor with internally and externally cooled annular fuel. The fuel assembly for annular case with 8 × 8 arrays is considered for annular pin configuration. The considered nanofluid is a mixture composed of water and particles of Al₂O₃ with various volume percentages. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Moreover, the temperature distributions of the fuel, clad and coolant are described for water/Al₂O₃ nanofluid in solid fuel and annular fuel. It is observed that as the concentration of Al₂O₃ nanoparticles increases, due to higher heat transfer coefficient of Al₂O₃ nanofluid, the temperature of the coolant is increased and the central fuel temperature is reduced. Thus, it improves margin from peak fuel temperature to melting. Finally, it is illustrated the use of the annular fuel instead of solid fuel in core of the reactor, security and efficiency of the nuclear power plant will be increased.     

Improvement of molten core–concrete interaction model in debris spreading analysis module with consideration of concrete degradation by heat

Understanding the situation inside of the reactors at TEPCO's Fukushima Daiichi Nuclear Power Plant and planning of the methods for debris removal are important for decommissioning the reactors. A debris spreading analysis (DSA) module in the severe accident analysis code SAMPSON has been improved and verified to analyze composite phenomena of molten core (debris) spreading on a reactor containment floor and concrete erosion to the inside of the floor by molten core–concrete interaction (MCCI). The primary models in the DSA module were three-dimensional natural convection with simultaneous spreading, melting and solidification in an open space. In addition to these, the analysis capability has been improved to treat phenomena in a closed space, such as debris eroding laterally under concrete floors at the bottom of the sump pit which is done by an advanced method for boundary processing. A buffer cell for flow analysis, which is defined by a different array variable, is arranged in the same coordinates of the concrete cell (structure cell). Mass, momentum, and the advection term of energy between the debris melt cells and the buffer cells are solved. At the same instant, the heat transfer is calculated between the debris melt cells and the structure cells coexisting side by side with the buffer cells. In this study, technical knowledge regarding changes in physical properties due to thermal degradation of concrete was considered for the prediction of erosion rate, and the DSA module with the models noted above was verified by comparison with erosion data of the core–concrete interaction tests in the OECD/MCCI program. The calculated erosion depth, width, and erosion rate under the concrete floor showed good agreement with the test data and the analysis capability of the module was confirmed.