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ASME Conference Presenter Attendance Policy and Archival Proceedings

2018;():V003T00A001. doi:10.1115/ICONE26-NS3.
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This online compilation of papers from the 2018 26th International Conference on Nuclear Engineering (ICONE26) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Nuclear Fuel and Material, Reactor Physics, and Transport Theory

2018;():V003T02A001. doi:10.1115/ICONE26-81020.

Nuclear power units need to operate conditioned the lowest risk possible. Safety analysis must use paired models, combining probabilistic and deterministic methods. In this study, FRAPCON and FRAPTRAN codes were used to simulate an idealized test based on IFA-650 series, carried out within Halden program. Nuclear systems work to depend on uncertainty values that must be quantified and propagated. The sources of uncertainties can be divided among physical models, boundary conditions, and mechanical tolerances. Eight physical models that can be configured, such as thermal conductibility, and fission gas release. Mechanical tolerances introduced by fuel fabrication are deviations that must propagate throughout of the system. To measure the effects produced by uncertainties were used correlation coefficients between entry and exit. Uncertainties contained on input values are spread to measure the impact created on safety limits. The method adopted used 96 samples to achieve the 95% of probability and 95% of confidence level.

Commentary by Dr. Valentin Fuster
2018;():V003T02A002. doi:10.1115/ICONE26-81041.

The concept of multi-dimensional heterogeneous resonance integral tables is proposed. The new type of resonance integral is designed for different fuel pins appearing in one lattice with two extra dimension of optical radius and number density ratio in the fuel. Numerical results show that this treatment improves the accuracy of embedded self-shielding method on irregular lattices.

Topics: Resonance
Commentary by Dr. Valentin Fuster
2018;():V003T02A003. doi:10.1115/ICONE26-81055.

The dynamic rod worth measurement method, which is used widely for PWR with square geometry lattices, is applied on WWER type reactor with hexagonal geometry lattices. RTNP code is developed to calculate the static spatial factors (SSF) and the dynamic spatial factors (DSF) through simulating the 3D space-time neutron dynamics and the response of excore detector during the measurement process. The improved quasi-static method is used for temporal discretization. The time-dependent shape equation is then transferred to a fixed source problem through backward Euler formula. And Multi-group Monte Carlo method is used to solve this fixed source problem. Four tests on Tianwan nuclear power plant (NPP) units 1&2 have been done since 2015. The measured results agree well with the predicted values. It takes about half hour per bank and 5 hours to measure all banks worth. The conventional boron dilution was used for rod worth measurement of Tianwan NPP units 1&2. It took at least 2 hours per bank and produced lots of boron wastes. The boron dilution has been replaced with this method for Tianwan NPP units 1&2.

Commentary by Dr. Valentin Fuster
2018;():V003T02A004. doi:10.1115/ICONE26-81090.

The UK Department of Business, Energy and Industrial Strategy (BEIS) recently launched an R&D programme in Digital Reactor Design, incorporating the development of a Nuclear Virtual Engineering Capability with an integrated Modelling and Simulation programme. A key challenge of nuclear reactor design and analysis is the system complexity, which arises from a wide range of multi-physics phenomena being important across multiple length scales. This project constitutes the first step towards developing an integrated nuclear digital environment (INDE) linking together models across physical domains and incorporating real world data across all stages of the nuclear lifecycle. Simulation case studies will be developed within the INDE framework, delivering an enhanced modelling capability while ensuring the framework has immediate application. For these case studies have been specified that are relevant to design and operation phases for AGR and PWR type reactors. The AGR case considers the through-life structural performance of graphite bricks. This involves modelling of multi-scale, multi-physics phenomena in the support of reactor operations. The PWR case study is based on core multiphysics modelling, with potential relevance to operating and future PWRs, and in particular in the design of SMRs.

Topics: Simulation , Design , Modeling
Commentary by Dr. Valentin Fuster
2018;():V003T02A005. doi:10.1115/ICONE26-81140.

Monte Carlo (MC) burnup calculation method, implemented through coupling neutron transport and point depletion solvers, is widely used in design and analysis of nuclear reactor. Burnup calculation is generally solved by dividing reactor lifetime into steps and modeling geometry into numbers of burnup areas where neutron flux and one group effective cross sections are treated as constant during each burnup step. Such constant approximation for neutron flux and effective cross section will lead to obvious error unless using fairly short step. To yield accuracy and efficiency improvement, coupling schemes have been researched in series of MC codes.

In this study, four coupling schemes, beginning of step approximation, predictor-corrector methods by correcting nuclide density and flux-cross section as well as high order predictor-corrector with sub-step method were researched and implemented in RMC. Verification and comparison were performed by adopting assembly problem from VERA international benchmark. Results illustrate that high order coupled with sub-step method is with notable accuracy compared to beginning of step approximation and traditional predictor-corrector, especially for calculation in which step length is fairly long.

Commentary by Dr. Valentin Fuster
2018;():V003T02A006. doi:10.1115/ICONE26-81169.

The finite element method based on unstructured mesh has good geometry adaptability, it has been used to solve reactor physics problems, manual description of geometric modeling and meshing makes the current finite element code very complicated, it greatly restricts the application of this method in the numerical calculation of reactor physics. Using the CAD pre-processing software ICEM-CFD, three dimensional geometry is divided into tetrahedral or hexahedral meshes, two dimensional geometry is divided into triangular or quadrilateral meshes, the main code of neutron calculation for nuclear noise analysis based on finite element method is developed. The steady state parameters are calculated and tested through benchmark problem, the test results show that the code has the corresponding computing capabilities. Finally, the neutron noise spectrum is calculated for the 3D PWR benchmark problem published by IAEA, and the noise distribution under given frequency is given.

Commentary by Dr. Valentin Fuster
2018;():V003T02A007. doi:10.1115/ICONE26-81170.

This paper designs a measure and analysis system of nuclear noise frequency spectrum for narrow channel research reactor core. Based on high-speed synchronization DAQ which is PXI-1031, the hardware design has been carried out. By using Labview, the data acquisition and analysis code has been developed. In order to verify the function of nuclear noise frequency spectrum measure and analysis system, firstly, through using arbitrary waveform generator which is Fluke-282, two sine signals have been tested. Secondly, by using history operation datum, the prompt neutron decay constant has been measured and analyzed. The simulation experimental and operational datum results show that such measure and analysis system of nuclear noise frequency spectrum is successful and reliable design, which can meet the design requirements. Finally, the preliminary research work between helium bubble and its responding frequency for narrow channel core has just started to monitor safety operation or carry out fault diagnosis.

Commentary by Dr. Valentin Fuster
2018;():V003T02A008. doi:10.1115/ICONE26-81174.

To improve the anti-oxidation ability of graphite, Al2O3 and SiC as additives, a uniform SiC coating was prepared on the surface of HTR graphite spheres by the pack cementation and sintering process. The microstructure of SiC coating was characterized by XRD, Raman spectroscopy, scanning electron microscopy, and the effect of Si/C ratio, Al2O3 and SiC content in the starting powder, sintering process conditions on the microstructure of SiC coating was also analyzed. The results show that the SiC coating on the surface of spherical graphite sample can be obtained with Si/C ratio higher than 3: 1, and the SiC coatings with different raw materials possess different microstructure and phase constitutes. The SiC coating is a kind of clear porous microstructure, while the main composition of the coating is β-SiC with some α-SiC and unreacted graphite. It is found that the content of Al2O3 has a significant influence on the penetration into the SiC coating, while the SiC content has a great influence on the microstructure between SiC coating and the substrate. When Al2O3 content is 10%, the bonding between the SiC coating and the substrate is better. When SiC content is 20%, the SiC coatings obtained have good densification and less lamellar accumulation of on the surface of the coatings. The SiC coatings sintered in Ar gas at 1750 °C have larger thickness and porous porosity microstructure, and the maximum thickness, about 400 ∼ 600 μm, can be achieved with the 2h of sintering time. The obtained SiC coatings sintered in vacuum at 1750 °C have a plate-like dense microstructure, in which the main composition of the coating is free Si, α-SiC and β-SiC, and the transitional area between SiC coating and the substrate is not obvious.

Topics: Coatings , Graphite
Commentary by Dr. Valentin Fuster
2018;():V003T02A009. doi:10.1115/ICONE26-81196.

Modular boiling water reactor (MBWR) can be considered as a small sized economic simplified boiling water reactor (ESBWR). It has the advantage of easier fabrication, transportation and construction. In this paper, a 65MWe MBWR core was designed with natural circulation, passive safety features, high power density and an 18 months fuel cycle. The MBWR core consists of 104 fuel assemblies with 4.6 w/o U-235, the assemblies were divided into 3 batches based on the depletion level, the batches shuffled at the end of each cycle. The core converged to equilibrium after 8 fuel cycles. A steady-state equilibrium fuel cycle depletion analysis was performed over a 540 day cycle using the HELIOS and PARCS software. The control blades insertion patterns were chosen to minimize axial and radial power peaking and provide uniform burnup throughout the cycle. At the end of the equilibrium cycle, 16% of total control blade worth remained inserted and the average assembly burnup is 21.318 GWd/MTHM. Thermal hydraulic analysis was also performed to insure the core’s safety feature.

Commentary by Dr. Valentin Fuster
2018;():V003T02A010. doi:10.1115/ICONE26-81205.

The point kinetics is very important to the safety of the reactor operation. However, these equations are stiff and usually solved with very small time step. These equations are solved by Revisionist integral deferred correction (RIDC), which is a parallel time integration method. RIDC is a highly accurate method, and it reduces the error by iteration. Based on C++ and MPI, a four-core fourth-order RIDC is implemented and tested by several cases, such as step, ramp, and sinusoidal reactivity insertion. Compared with other methods, the time step of RIDC in the step reactivity insertion case is smaller, but it’s larger in the case of the sinusoidal reactivity insertion. RIDC can keep high accuracy while the time step is appropriately large. The numerical results also show that the speed-up ratio can achieve 2 when 4 processors are used.

Commentary by Dr. Valentin Fuster
2018;():V003T02A011. doi:10.1115/ICONE26-81213.

Neutron kinetics plays an important role in reactor safety and analysis. The backward Euler method is the most widely used time integration method in the calculation of space-dependent nuclear reactor kinetics. Diagonally Implicit Runge-Kutta (DIRK) method owns high accuracy and excellent stability and it could be applied to the neutron kinetics for hexagonal-z geometry application. As solving the neutron kinetics equations is very time-consuming and the number of available cores continues to increase with parallel architectures evolving, parallel algorithms need to be designed to utilize the available resources effectively. However, it is difficult to parallel in time axis since the later moment is strongly dependent on the previous moment. In this paper, the Parareal method which is a time parallel method and implemented by MPI in the processor level is studied in the hexagonal-z geometry with the help of DIRK method. In order to make good use of the parallelism, a parallel strategy in the space direction is also used. In the coarse nodal method, many same operations are finished in the nodes and these operations could be parallel by OpenMP in the thread level since they are independent. Several transient cases are used to validate this method. The results show that the Parareal method gets a fast-convergent speed such as only 2∼3 iterations are needed to convergent. This space-time parallel method could reduce the cost time compared to the sequential method.

Commentary by Dr. Valentin Fuster
2018;():V003T02A012. doi:10.1115/ICONE26-81237.

ThO2 has been considered as a possible replacement for UO2 fuel for future generation of nuclear reactors, and thorium-based mixed oxide (Th-MOX) fuel performance in a light water reactor was investigated due to better neutronics properties and proliferation resistance compared to conventional UO2 fuel. In this study, the thermal, mechanical properties of Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel were reviewed with updated properties and compared with UO2 fuel, and the corresponding fuel performance in a light water reactor under normal operation conditions were also analyzed and compared by using CAMPUS code. The Th0.923U0.077O2 fuel were found to decrease the fuel centerline temperature, while Th0.923Pu0.077O2 fuel was found to have a bit higher fuel centerline temperature than UO2 fuel at the beginning of fuel burnup, and then much lower fuel centerline than UO2 fuel at high fuel burnup. The Th0.923U0.077O2 fuel was found to have lowest fuel centerline temperature, fission gas release and plenum pressure. While the Th0.923Pu0.077O2 fuel was found to have earliest gap closure time with much less fission gas release and much lower plenum pressure compared to UO2 fuel. So the fuel performance could be expected to be improved by applying Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel.

Commentary by Dr. Valentin Fuster
2018;():V003T02A013. doi:10.1115/ICONE26-81295.

Exposed to the high radiation environment, the PWR fuel rod will be swelling. The deformation is quite different azimuthally which caused by the different subchannels coolant temperature and some manufacture issues. And the asymmetry deformation will have a great effect on fuel rod safety for the increasing pellet center temperature and FCMI due to the change of gas gap. An accurate prediction on fuel strain which is associated with the temperature distribution closely should be done to assure the safety of the fuel rod. The conventional finite difference method has deficiency on treating with the deformation and the finite element method (FEM) could satisfy the requirement.

Based on the principle of FEM, the model of thermal conduction, stress-strain analysis is established. The two dimensional model is chosen considering the characteristic of fuel. The analysis code SSTFEM is developed and verified using the experimental data from the opening literature and results of authorized code. After that the code is used to analyze the effect of the asymmetry deformation which is caused by the asymmetry coolant temperature distribution in adjacent subchannels and the manufacture eccentricity between the fuel rod and the cladding. The effect law of deformation and the stress is obtained after a systematic analysis.

Commentary by Dr. Valentin Fuster
2018;():V003T02A014. doi:10.1115/ICONE26-81316.

The efficient solution of the neutron diffusion equation for large scale whole core calculations is of paramount importance; especially if the detailed pin-level power distribution and reaction rates are required. For heterogeneous whole core calculations finite element based techniques have been one approach to modelling the detailed pin geometry in whole core calculations. A new approach, pioneered in the last few years, is isogeometric analysis (IGA) methods which enable the exact geometry of fuel pins to be modelled. In order to efficiently solve elliptic partial differential equations (PDEs), such as the neutron diffusion equation, typically multi-grid or multi-level iterative solution techniques are used such as the algebraic multi-grid method. However, using IGA methods it is possible to develop true geometric multi-grid techniques which are potentially much more efficient than the standard algebraic multi-grid methods. In this paper we explore the use of IGA methods to develop a scalable, multi-level, iterative algorithm which is then used to solve the neutron diffusion equation over several geometries. This multilevel solution algorithm utilises a single patch multi-grid framework suggested by Hofreither and Takacs that takes advantage of the tensor product construction to provide scalability with respect to spatial, and polynomial refinement. Furthermore, a two-level balancing Neumann-Neumann solver is used to extend the solver to multiple patches in a scalable way. It is seen that the number of iterations required depends on the mapping between the unit square and the physical geometry, as well as the material coefficients.

Commentary by Dr. Valentin Fuster
2018;():V003T02A015. doi:10.1115/ICONE26-81317.

In this paper the application of the virtual element method (VEM) to the multigroup, neutron diffusion equation will be presented. The VEM is a recently developed Bubnov-Galerkin spatial discretisation method based largely on the mimetic finite difference method (MFD) that preserves the properties of the underlying vector operators. It can discretise elliptic partial differential equations (PDEs) on arbitrary polygonal/polyhedral meshes with arbitrary order and regularity. Deterministic, geometry conforming methods, used to solve multigroup, neutron diffusion, nuclear reactor physics problems have historically used the finite element method (FEM). However, FEM requires high-quality meshes with few highly distorted elements (elements with a poor aspect ratio) or it may may experience convergence problems. The process of creating high quality meshes, even with automated mesh generation algorithms, such as the advancing front and Delaunay methods, is often very time consuming. For these reasons VEM is being studied in this paper as a possible alternative to FEM in the numerical solution of neutron diffusion problems in nuclear reactor physics. A C5G7 UOX pincell problem is presented to demonstrate the application of VEM to mutligroup diffusion problem. The method of manufactured solutions (MMS) is used to determine the order of convergence.

Commentary by Dr. Valentin Fuster
2018;():V003T02A016. doi:10.1115/ICONE26-81322.

Until recently, interior penalty methods have been applied to elliptic operators using an approach based on the mass matrix of finite elements that possess a constant Jacobian. In the case of isogeometric analysis, such an approach would not always guarantee coercivity of the bilinear form and thus numerical stability of the solution. In this paper, optimal interior penalty parameters [1] are described and applied to the symmetric interior penalty scheme of the discontinuous Galerkin isogeometric spatial discretisation of the neutron diffusion equation. The numerical accuracy of the proposed method is compared against a standard continuous Bubnov-Galerkin isogeometric spatial discretisation of the neutron diffusion equation. Numerical consistency and order of error-convergence is verified by means of the method of manufactured solutions. Numerical results are also presented for a two-dimensional pin-cell test case, based upon the OECD/NEA C5G7 quarter core MOX fuel assembly benchmark for nuclear reactor physics parameters of interest.

Commentary by Dr. Valentin Fuster
2018;():V003T02A017. doi:10.1115/ICONE26-81356.

To solve three-dimensional kinetics problems, a high order nodal expansion method for hexagonal-z geometry (HONEM) and a Runge-Kutta (RK) method are respectively adopted to deal with the spatial and temporal problem. In the HONEM, 1D partially-integrated flux are approximated by using four order polynomial. The two order polynomial is adopted to the approximation of partially-integrated leakages. The Runge-Kutta method is adopted as a tool for dispersing the time term of 3D kinetics equation. A flux weighting method (FWM) is used for obtaining homogenized cross sections of mix node. The three-dimensional hexagonal kinetics code has been developed based on this method and tested with two benchmark problems of VVER which are the control rod ejection without any feedback and with simple adiabatic Doppler feedback. The results calculated by this code agree well with the reference results and the code is validated.

Topics: Neutrons , Geometry
Commentary by Dr. Valentin Fuster
2018;():V003T02A018. doi:10.1115/ICONE26-81384.

Taking the large commercial pressurized water reactor and its mature fuel assembly as reference, this paper has analyzed economic performance of two accident tolerant fuel (ATF) designs based on once-through fuel cycle. The results show that the fuel cycle costs of both AT F designs have grown due to application of BeO powder, which is expensive. In order to reach the same electric cost as that of the referred fuel assembly, burn-up of these two AT F designs should be enhanced to 51323MWd/tU and 52054MWd/tU respectively.

Commentary by Dr. Valentin Fuster
2018;():V003T02A019. doi:10.1115/ICONE26-81392.

Accelerator used for Boron Neutron Capture Therapy (BNCT) is the development trend of cancer treatment in the future. Neutron generator is a neutron source with compact structure, easy to operate, and lower price in accelerator (the structure of neutron tube is more compact). It has a high feasibility of establishing in the hospital when compared with other types of accelerator. At present, the D-D reaction has higher neutron yield than other reactions, more easy to get D material. Therefore, the epithermal neutron beam based on D-D neutron generator is studied for BNCT usage.

First, the calculation model is established by MCNP program, including the neutron source model, the geometric model of the irradiation device. The specific neutron energy spectrum and angular distribution of the D-D reaction are theoretically analyzed when establishing the D-D reaction neutron source model. The basic structure models of typical irradiation devices are designed, and then the optimal suitable moderator material for the irradiation device is studied from the neutron reaction cross section, the combination of iron and Fluental materials is optimum.

The irradiation device with multi D-D neutron tubes’ combination as BNCT neutron source is designed. It can be concluded by study that the parameters at the beam exit are epithermal neutron flux density 2.01 × 108 n/cm2 · s, fast neutron contamination 1.33 × 10−11 Gy · cm2/n, γ contamination 5.79 × 10−17 Gy · cm2/n, for the combination with 6 neutron tubes. The result can meet IAEA’s requirements for BNCT epithermal neutron beam quality.

Commentary by Dr. Valentin Fuster
2018;():V003T02A020. doi:10.1115/ICONE26-81468.

The thimble tube, which is made of Zircaly-4, is one of the main components of a PWR fuel assembly. The thimble tube has an important role as a structural member of the skeleton. Another role of the thimble tube is to guide a rod cluster control assembly (RCCA) for insertion during the reactor operation, and the function has to be assured not only in normal operation but in a seismic event.

In a horizontal seismic event, the fuel assembly vibrates laterally, which gives bending moment to the thimble tube. In addition, axial compressive force acts on the thimble tube in a vertical seismic event.

The integrity of the thimble tube has to be maintained while this force and moment act. Mitsubishi has confirmed by the elastic stress analysis that the stress of the thimble tube is lower than the limit value requested for the seismic event. The stress evaluation method is based on the ASME code.

The ASME code also describes the limit analysis which is available when the predicted stress is beyond elastic region of the material. In the analysis, the material is assumed to be elastic-perfectly plastic, and the maximum load that the structure can carry is calculated. For the reason mentioned above, the allowable limit of the thimble tube should be determined as a function between the force and the moment. We are planning to examine the allowable limit experimentally.

As a step before testing, an analytical approach for the limit is discussed in this paper. Firstly, the allowable limit is calculated by a beam model assuming elastic-perfectly plastic material, based on the ASME code.

Secondly, a 3D model analysis with elastic-plastic material is performed to predict the practical strength. Based on the comparison with the analysis using elastic-perfectly plastic material, ASME based limit is considerably conservative compared with the one with the actual stress-strain curve. Conversely, this means there is enough room to rationalize the allowable limit.

As the future work, the experiment will be conducted to obtain the practical limit of the thimble tube and to verify the analysis results.

Commentary by Dr. Valentin Fuster
2018;():V003T02A021. doi:10.1115/ICONE26-81474.

In this paper, preliminary neutron physical properties of ceramic fast reactor (CFR) are simulated and analyzed. The CFR core consists of ceramic materials, including nuclear fuels, coolants, structural materials, reflective and absorption materials. These ceramics improve inherent safety levels substantially, increase breeding performance, and enhance the power-generation efficiency. The CFR has the potential to operate and breed more than 30 years. The performance of the CFR was simulated focusing on neutron-related effects. The parameters discussed contain fast neutron energy spectrum, the ideal effective multiplication-factor, nuclides mass changes, breeding performance, operation mode, etc. Furthermore, the strengths of the proposed reactor system are discussed. In the future nuclear energy system, CFR may be one of the existing alternative novel reactor type.

Commentary by Dr. Valentin Fuster
2018;():V003T02A022. doi:10.1115/ICONE26-81507.

The 2D/1D fusion method (2D/1D method) is becoming a popular transport method for whole-core calculations, which reduces the group condense and assembly homogenization approximations in the conventional two-step reactor physics calculations. In most 2D/1D codes, a pin is chosen as a 1D calculation domain, which assumes that the axial leakage of the pin is flat on top/bottom surfaces. Similar to the axial leakage, the radial leakage of every 2D plane also introduces several approximations along axial direction for the 1D calculation. In this paper a 2D/1D fusion code is developed, while a leakage reconstruction method is proposed and applied. In this 2D/1D fusion code, MOC is applied to the radial 2D calculation and the Sn diamond difference method is used for the axial 1D calculation. Numerical results indicate that the 2D/1D fusion code developed in this paper is precise in three-dimensional transport calculation and show the performance of this leakage reconstruction method especially when the leakage term is significant.

Topics: Leakage
Commentary by Dr. Valentin Fuster
2018;():V003T02A023. doi:10.1115/ICONE26-81509.

Flow elastic stability and vortex shedding were two important mechanisms of the flow induced vibration analysis. Due to the influence of manufacturing process, transportation and irradiation, the clamping action of grid on fuel rods may be invalid. Taking one fuel assemblies as an example, the effects of clamping failure on the natural frequencies, mode shapes, flow elastic stability and vortex shedding were studied. The results show that the effect of the rigid convex support failure on the natural frequency was directly related to the mode shape. The effect of the grid rigid convex failure near the nodes with larger amplitude on the natural frequency was obvious. The velocity of flow at the top and bottom of the fuel rods were larger and the size was comparable, this induced that the rigid convex failure of the top and bottom grids had a significant effect on the ratio of flow elastic stability and the vortex shedding.

Commentary by Dr. Valentin Fuster
2018;():V003T02A024. doi:10.1115/ICONE26-81516.

To process the evaluated nuclear data file (ENDF) libraries and generate the cross section data library for neutronics calculations, a new nuclear data processing system NECP-Atlas was developed by Nuclear Engineering Computational Physics Lab. of Xi'an Jiaotong University. Meanwhile, some flaws of the current widely used nuclear data processing systems were made up.

Some new methods and techniques were proposed and integrated into NECP-Atlas. NECP-Atlas could process ENDF and generate point-wise evaluated nuclear data file (PENDF) and the multigroup cross section data library in WIMS-D format. Verification of NECP-Atlas was carried out by comparing the keff values for WLUP benchmark cases and benchmark experiments in the ICSBEP handbook using cross section data libraries processed by NECP-Atlas with those by NJOY2016. The results showed that NECP-Atlas processes the ENDF correctly and generates more reliable cross section data libraries.

Commentary by Dr. Valentin Fuster
2018;():V003T02A025. doi:10.1115/ICONE26-81540.

Two kinds of new cladding material for pressurized water reactor were developed by China Nuclear Power Research and Technology Institute (CNPRI) which named CZ1 and CZ2. The cladding tubes of CZ alloys used in the study were fabricated by different final annealing temperature in the range of 450 °C to 600 °C.In order to investigate effect of final annealing temperature on creep property of CZ alloys. The axial creep tests were carried out on specimens of CZ-SRA and CZ-RXA At the temperature of 375 °C with different applied stress levels for 1000 hours. The experimental results showed that the axial creep resistance of CZ-RXA is superior to CZ-SRA at lower stress and is inferior to CZ-SRA at higher stress. The mechanism was discussed.

Commentary by Dr. Valentin Fuster
2018;():V003T02A026. doi:10.1115/ICONE26-81541.

Breakaway oxidation is a potential fuel failure mechanism during a loss-of-coolant accident (LOCA), especially small-break LOCA. Two kinds of new advanced zirconium alloy, CZ1 and CZ2, were developed by China Nuclear Power Research and Technology Institute of China General Nuclear Power Group. The breakaway oxidation behavior of CZ1 and CZ2 was studied. The outer surface of all samples was examined visually and photographed. After 2730s oxidation in steam, the outer surface of CZ1 alloy sample and CZ2 alloy sample remained lustrous-black. The outer surface of CZ1 sample oxidized in steam at 1000 °C for 4181s was grey, but under the same experimental conditions the outer surface of CZ2 sample was still lustrous black. The hydrogen pickup content of different oxidation time was measured. The samples with grey appearance showed significant hydrogen pickup. The microstructure was observed by optical microscope. Evolution of oxide structure was described, and the mechanism was discussed.

Topics: Alloys , oxidation
Commentary by Dr. Valentin Fuster
2018;():V003T02A027. doi:10.1115/ICONE26-81574.

The FRED code is an in-house tool developed at the Paul Scherrer Institut for the so-called 1.5-D nuclear fuel performance analysis. In order to extend its field of application, this code has been re-implemented as a class of the OpenFOAM numerical library. A first objective of this re-implementation is to provide this tool with the parallel scalability necessary for full-core analyses. In addition, the use of OpenFOAM as base library allows for a straightforward interface with the standard Open-FOAM CFD solvers, as well as with the several OpenFOAM-based applications developed by the nuclear engineering community. In this paper, the newly developed FRED-based Open-FOAM class has been integrated in the GeN-Foam multi-physics code mainly developed at the École polytechnique fédérale de Lausanne and at the Paul Scherrer Institut. The paper presents the details of both the re-implementation of the FRED code and of its integration in GeN-Foam. The performances and parallel scalability of the tool are preliminary investigated and an example of application is provided by performing a full-core multi-physics analysis of the European Sodium Fast Reactor.

Topics: Physics , Fuels
Commentary by Dr. Valentin Fuster
2018;():V003T02A028. doi:10.1115/ICONE26-81595.

In presence of a weak neutron source, the initial growth of neutron population in a supercritical system exhibits a significant stochastic feature, both initiation and burst waiting times are uncertain. As a result, the energy released during criticality excursions is stochastic, obeying a probability distribution. When criticality accidents and pulsed reactor experiments are analyzed, it is important to estimate this kind of stochastic feature, including assessing the initiation probability and then the fission energy probability distribution. Thus a Monte Carlo direct simulation method has been proposed and the corresponding code MES has been developed. By taking random factors during criticality excursions into account in dynamic Monte Carlo simulations, this method is capable of simulating the whole process from source injection to exponential growth of the neutron population, and finally to extinction of the neutron pulse. A set of static initiation probability problems and a figurative criticality excursion problem have been applied to validate this method and MES. Results demonstrate that with the proposed method MES is able to simulate stochastic transient neutron fields in multiplying systems during criticality excursions.

Commentary by Dr. Valentin Fuster
2018;():V003T02A029. doi:10.1115/ICONE26-81596.

In order to ensure the safety of fuel rods in nuclear reactor, it is necessary to consider the condition of the power ramp during reactor operation, which may cause breakage risk of fuel rod at the time of pellet-cladding interaction (PCI) appearing. To analyze this phenomenon and reduce the risk, a performance analysis model for fuel rod is developed to carry out the steady state and transient simulation using commercial software COSMOL. The full model has three main computational models, which are heat transfer model, mechanical model and internal pressure model. The calculation results show that the stress and strain of NHR200-II fuel rods has enough margins during the power ramp process, which indicates that the risks of fuel rod damage are very low.

Commentary by Dr. Valentin Fuster
2018;():V003T02A030. doi:10.1115/ICONE26-81621.

Spent nuclear fuel (SNF) integrity evaluation related to its handling and transportation for mid-/long-term dry storage is a regulatory requirement. Especially, a drop event is the most fatal failure mode among regulatory conditions. For SNF drop accidents, it is required that the mechanical integrity of the SNF be evaluated using test results or analytic methodologies. The SNF mechanical test, however, takes much time and cost, and there are safety issues related to the release of radioactive materials. Thus, finite element analysis is used as an alternative to the experimental test method to solve this problem. In this study, a three dimensional (3D) finite element model was developed using ABAQUS software to simulate the structural behaviors of a fresh fuel assembly (FA) prior to applying SNF properties because of a lack of SNF test results. Static and dynamic mechanical behaviors were simulated with this model and compared with the fresh FA test results. The analysis results are in good agreement with the test results. Therefore, the analysis model consistent with the test results will be applied to the evaluation of the FA drop integrity reflecting the specific SNF characteristics.

Commentary by Dr. Valentin Fuster
2018;():V003T02A031. doi:10.1115/ICONE26-81665.

Delayed hydride cracking (DHC) is the result of a mechanism of crack initiation and slow propagation. In DHC, hydrogen diffusion in the metal is required. Gradients of concentration, temperature, and stress are all important factors in controlling diffusivity. The classic theory of DHC still has potential to be modified. In this study, a calculation formula for DHC SIF threshold is established with consideration of the temperature history, the temperature field and the stress field induced by the temperature gradient and the external mechanical loading. Moreover, the temperature gradient on crack surfaces has been considered in the model.

Commentary by Dr. Valentin Fuster
2018;():V003T02A032. doi:10.1115/ICONE26-81709.

Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor types and research objections. But usually it’s difficult to converge well enough to deal with deep penetration problems, even though there are a number of variance reduction techniques. Based on the advantages and disadvantages of Monte Carlo and Deterministic method, we proposed a coupled neutron transport calculation method for deep penetration. It combines advantages of these two methods. Firstly, we use Monte Carlo code to finish fine modeling and do the whole reactor core calculation. Domestically developed Monte Carlo code JMCT is used to do the neutron transport calculation. Then homogenized group constants in each mesh are calculated from JMCT output by a self-developed script. Afterwards, we do the whole reactor calculation with deterministic theory code TORT. It directly uses group constants generated by Monte Carlo code. Finally, we can get the deep penetration calculation results from TORT output. Verification is carried out by comparing the group constants of benchmark problem, and by comparing keff calculated by this method with continuous energy Monte Carlo method. Benchmark calculation is conducted with OECD/NEA slab benchmark problem. The comparison shows that group constants generated by this study are in good agreement with results from published references. Then above group constants are applied to 3-dimensional discrete ordinates deterministic theory transport code TORT. But keff calculated by TORT is a little lower than that calculated by Monte Carlo code JMCT. To minimize other influence factors, different Sn/Pn order, and different mesh size in TORT has been tried. Unfortunately the keff difference between these two methods remains. Even though the keff results in this benchmark are less than keff calculated by continuous energy MC method, Benchmark results show that all the group constants generated by this method are in good agreement with existing references. So it can be expected that after further verification and validation, this coupled method can be effectively applied to the deep penetration problem in such kind of research reactors.

Commentary by Dr. Valentin Fuster
2018;():V003T02A033. doi:10.1115/ICONE26-81711.

In recent years there has been an increasing demand from nuclear research, industry, safety, and regulation bodies for best estimate predictions of Light Water Reactors (LWR’s) performances to be provided with their confidence bounds. From a neutronic point of view, among the different sources of uncertainty the main challenge is represented by the one related to the accuracy of the nuclear data libraries used in the transport calculations. The assessment of nuclear data uncertainties and their impact on the main reactor parameters plays a fundamental role not only for criticality safety but also in burnup analyses. In facts, the accurate prediction of nuclear parameters in burnup calculations strongly affects the management of spent nuclear fuel, the core design, as well as the economy and safety of nuclear reactors. In this paper a study related to the impact of the nuclear data uncertainties on the evolution in time of the criticality and the nuclide concentrations in burnup calculations is presented. The analysis has been performed by using a statistical sampling methodology in which all the uncertain parameters are handled as random dependent variables by a sampling procedure. The probability distributions of the uncertain input parameters are used to generate random variations of these input quantities starting from a covariance library in a 56-group energy structure. Calculations have been performed by means of the SCALE 6.2.2 code and ENDF/B-VII.1 nuclear data. The method has been tested on a PWR pin cell model representative of the TMI-1 15 × 15 assembly as defined in an international benchmark exercise. In the first part of the paper the methodology and the neutronics modelling of the problem are presented.

Commentary by Dr. Valentin Fuster
2018;():V003T02A034. doi:10.1115/ICONE26-81923.

Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This study performed SA analyses considering high temperature chemical reaction characteristics of SiC by using SA analysis code “MAAP”, and thermal hydraulics analysis code “TRAC Toshiba version (TRAC)”, and compared the difference between SiC and Zry.

Both codes originally have no model of oxidation reaction for SiC. Hence, a new model for SiC in addition to the current model for Zry was incorporated into “MAAP”. On the other hand, “TRAC” adjusted reaction rate by changing oxidation reaction coefficients in the current Zry oxidation reaction models such as Baker-Just and Cathcart correlations in order to simulate SiC-water/steam reaction.

In analysis using “MAAP”, seven accident sequences from representative Probabilistic Risk Assessment ones were selected to evaluate the difference of SA behavior between two materials. As a result, in the case of replacing current Zry of fuel claddings and channel boxes into SiC, an amount of hydrogen generation reduced to about 1/6 than the case of Zry. In addition to that, in the case of replacing SUS structures in the reactor core into SiC, an amount of hydrogen generation moreover reduced to about 1/6 than the above result, which means just about 2% of an amount in the original case.

On the other hand, in analysis using “TRAC”, the accident sequence for unit 3 of 1F (1F3) was selected, and reaction rate in the oxidation reaction model was examined as parameter. In the case of 1.0 time of the reaction rate, which means an original reaction rate, maximum fuel cladding temperature exceeded 2000K in 50 hour after reactor scram. However, using the reaction rate below 0.01 to the original one, the fuel cladding temperature didn’t exceed 1,600K.

Topics: Fuels , Accidents
Commentary by Dr. Valentin Fuster
2018;():V003T02A035. doi:10.1115/ICONE26-81982.

As people pay more attention to nuclear safety analysis, sensitivity and uncertainty analysis has become a research hotspot. In our previous research, we had developed an integrated, built-in stochastic sampling module in the Reactor Monte Carlo code RMC [1]. Using this module, we can perform nuclear data uncertainty analysis. But at that time the uncertainty of fission spectrum was not considered. So, in this work, the capability of computing the uncertainty of keff induced by the uncertainty of fission spectrum, including tabular data form and formula form, is implemented in RMC code based on the stochastic sampling method. The algorithms and capability of computing keff uncertainty induced by uncertainty of fission spectrum in RMC are verified by comparison with the results calculated by the first order uncertainty quantification method [2].

Commentary by Dr. Valentin Fuster
2018;():V003T02A036. doi:10.1115/ICONE26-82008.

This paper presents a fast sub-grid scale (SGS) finite element method for the first order neutron transport equation. The spherical harmonics method is adopted for the angular discretization. The sub-grid scale discretization embeds discontinuous component in each element to provide a stabilization term for the continuous finite element formulation. Traditional SGS method uses Riemann decomposition and vacuum boundary assumption to decouple the discontinuous component. Here we propose a new method to perform the decoupling based on the assumption that the convection term of the discontinuous component is proportional to the residual of angular flux in each element. The computing costs for the establishment of the coefficient matrix of discontinuous component are reduced to O(1) from O(n3). Further more, the computing costs for the inversion of the coefficient matrix are reduced to O(n) from O(n3) by applying mass lumping technique. Numerical results show that the new method is not only more efficient but also yields more accurate solution than traditional sub-grid scale method.

Commentary by Dr. Valentin Fuster
2018;():V003T02A037. doi:10.1115/ICONE26-82009.

The safety of radioactive material transport is regulated by graded approach in China. National requirements on container design, manufacture and shipment are different for the three categories material. Packages of different categories shall be experienced different test to demonstrate their ability to withstand transport normal transport conditions and accident conditions depending on their contents. China Institute for Radiation Protection (CIRP) established free drop test facilities, thermal test facilities, water immersion test facilities and relevant measurement means. CIRP has carried out tests of 18 radioactive material transport packages, and accumulate a lot of experiences about the limit unit analysis, test scheme, test process, monitoring and data treatment. Probabilistic safety assessment method and traffic accident statistics in China are carried out in the recent years by CIRP.

Commentary by Dr. Valentin Fuster
2018;():V003T02A038. doi:10.1115/ICONE26-82098.

The main objective of this work is to produce an optimal modeling for our aged Planar-HPGe detector using Monte Carlo method (MC). That optimization included the analysis of the germanium dead (inactive) layer thickness for our old detection system (planar-HPGe detector). DL is one of the important parameters needed in order to obtain the smallest discrepancy between simulated and experimental measurements of detector efficiency. Also, precise determination of 235U enrichment for UO2 samples which is necessary for purposes of nuclear materials verification in the field of nuclear safeguards.

The thickness of Germanium dead layer (DL) can be vary by time as it is not well known due to the existence of a transition zone where photons are strongly attenuated and absorbed, that cannot contribute to the total photon energy absorption which causes a significant decrease in efficiency. Therefore, using data provided by manufacturers since long years (manufacture date) in the detector simulation model is not convenient. As a result, some strong discrepancies appear between measured and simulated efficiency, in addition to that non-accurate results for 235U enrichment determination. The Monte Carlo method applied to overcome this difficulty was to vary the thickness of dead layer step by step in simulation, a good agreement (minimum deviation) between estimated and experimental efficiency was reached when a suitable germanium dead layer thickness was chosen. Calculations and measurements were performed for radioactive nuclear material samples in the form of UO2 powder with different sizes and enrichments at different locations, under different gamma-lines emitted after a-decay of the 235U nuclei. Results indicated that a good agreement between simulated and measured efficiencies is obtained using a value for the germanium dead layer thickness approximately (2.45 mm) six in comparison with (0.389 mm) provided by the detector manufacturer.

Topics: Sensors , Simulation
Commentary by Dr. Valentin Fuster
2018;():V003T02A039. doi:10.1115/ICONE26-82101.

The 235U enrichment is one of the most important characteristics of nuclear materials for nuclear safeguards purposes. The multi-group γ-ray analysis method for uranium (MGAU) is an important non-destructive gamma spectroscopy method for 235U enrichment determination. Using that method, the typical measurement bias is below 3% for uranium material with abundance from 0.3 to 93 %. However, it is not applicable for the samples with thick container or without isotopic decay equilibrium. In this work, the enrichment meter method was studied with two uranium dioxide samples (235U abundance 0.71 % and 3.167 %). The nuclear materials spectra were measured using a planar high-purity germanium detector. Based on the specific gamma peak (185.71 keV) of relative high intensity, this traditional enrichment meter approach gives measurement bias more than 10 %. Thus, this work represents two objects: (1) an alternative approach which was investigated, where the calibration is performed through Monte Carlo simulation (MCNP5) instead of experiment in advance, as the measurement bias was reduced to be around 5 %. Thus, to use this approach, one should have the sample details, such as dimensions, chemical composition and container. (2) The influence of the container wall thickness on the measurement accuracy by Monte Carlo simulation. So, if the container wall thickness is not modeled correctly the measurement accuracy is influenced, which is investigated by simulation.

Topics: Uranium
Commentary by Dr. Valentin Fuster
2018;():V003T02A040. doi:10.1115/ICONE26-82125.

The pressure and coolant temperature of Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY200) is significantly lower than PWR of the NPP, the core design and analysis were completed according to the design parameters and features of HAPPY200. The fuel assembly and its feature was firstly designed and studied based on the investigation of different types of fuel assemblies. Then the core configuration was studied and optimized according to the design parameters of HAPPY200; Eventually, neutronics calculation of the core was performed and key parameters were obtained including cycle length, power distribution, control rod worth, reactivity coefficients and etc. The study shows that with the core design HAPPY200 can be operated for 18 months in full power and reactivity control system can maintain criticality of the core in the full cycle. Due to the non-soluble boron design of the reactivity control scheme, moderator temperature coefficient and isothermal temperature coefficient are both negative, the Doppler temperature coefficients and power coefficients in different phase of the lifetime and in different power levels are also negative, therefore, the reactivity safety of the reactor core can be ensured.

Topics: Design
Commentary by Dr. Valentin Fuster
2018;():V003T02A041. doi:10.1115/ICONE26-82132.

Radiation-induced damage and degradation to shielding and structure materials in nuclear reactors is one of the key limiting factors that affect the safety considerations and the lifetime. Neutron radiation damages materials mainly by exciting a number of Primary Knock-on Atom (PKAs). PKAs induce displacement cascades, causing microstructure changes and mechanical degradations in materials. Computer simulations are used to model this complex process. Knowing the PKA spectrum accurately is important because PKA spectrum and the input of computer simulation are coupled. In this work, we aim to obtain the PKA spectrum in Zr-based alloys with relatively low Zr concentration by using the Geant4 software. Some new functions were added by reprogramming in Geant4. We found that the energy spectra of PKA in Zr2Cu and Zr2Ni are mainly caused by Zr atom, and the shape of the average energy spectra are similar with pure Zr. The number of PKA distributions and the energy deposition in these two Zr-based metals are similar with pure Zr but different than those in pure Ni. These finding indicate that the metallic elements of Cu and Ni have small impact on PKA spectrum in Zr-based alloys, which illustrate that simplified simulation models are feasible when using computer for simulating. Moreover, it has great significant for the calculation of irradiation damage and the computer simulation for the process of collision cascade after one PKA is formed.

Commentary by Dr. Valentin Fuster
2018;():V003T02A042. doi:10.1115/ICONE26-82144.

The computing power available nowadays to the average Monte-Carlo-code user is sufficient to perform large-scale neutron transport simulations, such as full-core burnup or high-fidelity multiphysics. In practice however, software limitations in the majority of the available Monte Carlo codes result in a low efficiency when running in High Performance Computing (HPC) environments, the main issues being inadequate memory utilization and poor scalability. The traditional parallel processing scheme based of splitting particle histories among processes requires domain replication across nodes, and therefore the memory demand for each computing node does not scale, and a memory bottleneck appears for large-scale problems. The scalability of this approach usually limits the resources that can be used efficiently to a small number of nodes/processors. Consequently, massively parallel execution is not viable with particle-based parallelism, at least not by itself. In this work we propose a Spatial Domain Decomposition (SDD) approach to develop an efficient and scalable Monte Carlo neutron transport algorithm. Breaking down the geometry into subdomains, a distributed memory scheme can be used to reduce the in-node memory demand, allowing the simulation of large-scale memory-intensive problems. Additionally, with an efficient neutron tracking algorithm the overall speedup can be significantly improved.

Topics: Neutrons
Commentary by Dr. Valentin Fuster
2018;():V003T02A043. doi:10.1115/ICONE26-82185.

The adjoint neutron flux is vital in the analysis of reactor kinetics parameters and reactor transient events. Both deterministic and Monte Carlo methods have been developed for the adjoint neutron flux calculation on the basis of multi-group cross sections which may vary significantly among different types of reactors. The iterated fission probability (IFP) method is introduced to calculate the neutron importance which is able to represent the adjoint neutron flux in continuous energy problem and have been applied to the calculation of kinetics parameters. However, the adjoint neutron flux can’t be obtained directly and applied to both Monte Carlo transient event analysis and deterministic methods. In this study, a method based on IFP is studied and implemented in Monte Carlo code RMC. The multi-group adjont neutron flux can be obtained directly through the discretization of energy and space with the modification of fission neutrons through continuous energy Monte Carlo calculations. The obtained multi-group adjoint neutron flux can be used in both Monte Carlo transient analysis and deterministic methods.

Topics: Neutron flux
Commentary by Dr. Valentin Fuster
2018;():V003T02A044. doi:10.1115/ICONE26-82191.

Based on the tri-isotropic (TRISO) fuel technology of high temperature gas cooled reactor, a new fuel concept for improving the accident tolerance of light water reactor (LWRs) named inert matrix dispersion pellet (IMDP) was proposed. Through the silicon carbide matrix and embedded TRISO fuel particles, the safety of the nuclear fuel could be enhanced. Recently, dummy IMDPs were fabricated by China General Nuclear Power Corporation (CGN) and thermal conductivity was tested. According to the tested data, a FEA model using ABAQUS combined its secondary development function was developed and benchmarked. Several influence factors of the effective thermal conductivity (ETC) of the IMDP were studied by the FEA model, such as burn-up, TRISO packing fraction and temperature. The heat transfer behaviors of IMDP and UO2 under typical normal PWR operating condition were also studied.

Commentary by Dr. Valentin Fuster
2018;():V003T02A045. doi:10.1115/ICONE26-82196.

Global-local self-shielding calculation scheme is a new high-fidelity resonance calculation model proposed by NECP laboratory of Xi'an Jiaotong University. Neutron Current Method (NCM) is utilized for resonance calculation in the global aspect to obtain Dancoff factors. Then each fuel pin is transformed into individual 1D cylindrical problems by conserving Dancoff factors. The Pseudo-Resonant-Nuclide Subgroup Method (PRNSM) is used to conduct resonance calculation in the local aspect for each 1D cylindrical pin. Global-local self-shielding calculation scheme has been successfully implemented in high-fidelity numerical nuclear reactor physics code NECP-X. Verification results of global-local self-shielding calculation scheme showed good accuracy for UO2 fuels. The maximum relative error of microscopic absorption cross sections (XSs) for 238U in resonance range was 1.5% compared with MCNP5 [1]. AIC control rods serve as strong absorbers in reactor. Strong self-shielding phenomenon occurs when AIC control rods are inserted. Analysis was performed to determine the effects of AIC control rods on the accuracy of global-local self-shielding calculation scheme and the sources of error. Evaluation results showed that the main part of error was introduced by NCM and radius searching. The relative errors were larger than 10% in several resonance groups. Therefore, a supercell model is proposed to couple with global-local self-shielding calculation scheme to treat resonance calculation for AIC control rods in this paper. Numerical results show that this model improves the accuracy of the global-local self-shielding calculation scheme. The relative errors of microscopic absorption XSs for AIC in most resonance groups were decreased to less than 2%.

Topics: Rods
Commentary by Dr. Valentin Fuster
2018;():V003T02A046. doi:10.1115/ICONE26-82242.

Tristructural isotropic (TRISO) fuel particles are chosen as the major fuel type of High temperature gas cooled reactor (HTGR). The TRISO coated particle also acts as the first barrier for radioactivity retention. The performance of the TRISO coated particle has a significant influence on the safety of HTGR. A set of fuel performance analysis codes have been developed during the past decades. The main functions of these codes are conducting stress calculation and failure probability prediction. PANAMA is a widely used German version fuel performance analysis code, which simulates the mechanical performance of TRISO coated particle under normal and accident conditions. In this code, only a simple pressure vessel model is considered, which is insufficient in stress analysis and fuel failure rate prediction. Nowadays, efforts have been done to update the fuel performance model utilized in PANAMA code, and a new TRISO fuel performance analysis code, FFAT, is under developed. This paper describes the newly updated TRISO fuel performance model and presents some first results based on the updated model.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2018;():V003T02A047. doi:10.1115/ICONE26-82244.

Heat pipe-segmented thermoelectric module converters space reactor power system (HP-SMTCs SRPS) is a promising candidate for space nuclear power system. An examination was taken to discuss the criticality safety of HP-SMTCs reactor core in several accident conditions. In the original nuclear design, the case that reactor core is submerged in wet sand while the voids inside the core are filled with sea water and the BeO reflector are dismantled is defined as the worst status with the highest risk of supercritical. However, recent Monte Carlo transport calculation result shows that reactivity in the case that the core is submerged in water while the voids inside are empty is even higher than those cases with the voids full of sea water, which means that the reactor may encounter high risk of supercritical when some particular accidents occur. Detailed analysis about the neutron energy spectrum and absorption reaction rate is made in order to find out the potential reason of these unexpected results. According to the discussion about the criticality safety issues in some accidents, further evaluations may be necessary for the neutronics design of HP-SMTCs space reactors.

Commentary by Dr. Valentin Fuster
2018;():V003T02A048. doi:10.1115/ICONE26-82298.

The spacer grid is a key element of the fuel assembly used in the Pressurized water reactor. Due to its structural complexity, the analysis and the design of the spacer grid structure is difficult. This paper discusses the 5 × 5 cell size partial grid analysis including the detailed grid structural elements, through which, the impact force, the rebound velocity and the time history of acceleration and as well as other mechanical properties of grid under different initial impact velocity were obtained. This paper carried out the dynamic buckling criterion studies, and determined the dynamic buckling load of the 5 × 5 cell size partial spacer grid. Based on assuming the impact process is simple harmonic vibration, the method to determine the dynamic stiffness of the spacer grid was proposed. The experiments were also performed for the comparison with the analytical results. It is found that the analytical results are in good agreement with the experimental results. As a result, we can conclude that the analysis model including detailed grid elements is able to yield accurate analytical results.

Topics: Buckling
Commentary by Dr. Valentin Fuster
2018;():V003T02A049. doi:10.1115/ICONE26-82305.

Continuous improvement in Nuclear Industry Safety Standards and reactor designer’s and operator’s commercial goals lead to an increasing demand of fast running and highly accurate methodologies oriented to improve the prediction capabilities of main reactor’s parameters under steady state and transient situations.

An increasing effort has been observed during past years to develop high accurate multi-physics approach for nuclear reactor analysis, based both on the availability of advanced codes and the constant increase of computational resources with massive parallel architectures. As a result, several improvements have been observed in the implementation of coupled full 3-D Monte Carlo (MC) neutronic models for nuclear reactor cores, including not only the coupling to thermal-hydraulics but also fuel behavior codes. This approach has proved at concept level to be able to develop high accurate models that would allow to predict important safety and performance parameters of nuclear reactors with less conservativism.

Under this framework, the European Research and innovation project McSAFE is a coordinated effort started in September 2017 with the objective of moving MC stand-alone and coupled solution methodologies to become valuable tools for core design, safety analysis and industry like applications for LWRs of gen II and III.

In this work the Serpent 2 code, a high performance MC code developed by VTT, is used by the aim of performing the preliminary screening of capabilities, performance and limitations of such challenging objective.

As a result, simplified analysis are developed to identify full 3-D modeling computational requirements for typical LWR configurations, including burnup aspects. Potential bottlenecks and limitations are presented and discussed, providing foreseen alternatives and solutions for further code improvements.

Commentary by Dr. Valentin Fuster
2018;():V003T02A050. doi:10.1115/ICONE26-82352.

The isotope Xe-135 has a large thermal neutron absorption cross section and is considered to be the most important fission product. A very small amount of such neutron poison may significantly affect the reactor reactivity since they will absorb the neutrons from chain reaction, therefore monitoring their atomic density variation during reactor operation is extremely important. In a molten reactor system, Xe-135 is entrained in the liquid fuel and continuously circulates through the core where the neutron irradiation functions and the external core where only nuclei decay occurs, at the same time, an off-gas removal system operates for online removing Xe-135 through helium bubbling. These unique features of MSR complicate the Xe-135 dynamic behaviors, and the calculation method applied in the solid fuel reactor system is not suitable. From this point, we firstly analytically deduce the nuclei evolution law of Xe-135 in the flowing salt with an off-gas removal system functioning. A study of Xe-135 dynamic behavior with the core power change, shutdown, helium bubbling failure and startup then is conducted, and several valuable conclusions are obtained for MSR design.

Commentary by Dr. Valentin Fuster
2018;():V003T02A051. doi:10.1115/ICONE26-82381.

A new 3D fuel behavior solver is currently under collaborative development at the Laboratory for Reactor Physics and Systems Behaviour of the École Polytechnique Fédérale de Lausanne and at the Paul Scherrer Institut. The long term objective is to enable a more accurate simulation of inherently 3D safety-relevant phenomena which affect the performance of the nuclear fuel. The current implementation is a coupled three-dimensional heat conduction and linear elastic small strain solver, which models the effects of burnup- and temperature dependent material properties, swelling, relocation and gap conductance. The near future developments will include the introduction of a smeared pellet cracking model and of material inleasticities, such as creep and plasticity. After an overview of the theoretical background, equations and models behind the solver, this work focuses on the recent preliminary verification and validation efforts. The radial temperature and stress profiles predicted by the solver for the case of an infinitely long rod are compared against their analytical solution, allowing the verification of the thermo-mechanics equations and of the gap heat transfer model. Then, an axisymmetric model is created for 4 rods belonging to the Halden assembly IFA-432. These models are used to predict the fuel centerline temperature during power ramps recorded at the beginning of life, when the fuel rod performance is still not affected by more complex high burnup effects. Finally, the predictions are compared with the experimental measurements coming from the IFPE database. This first preliminary results allow a careful validation of the temperature-dependent material properties and of the gap conductance models.

Topics: Nuclear fuels
Commentary by Dr. Valentin Fuster
2018;():V003T02A052. doi:10.1115/ICONE26-82385.

Uncertainty quantification is an indispensable analysis for nuclear reactor simulation as it provides a rigorous approach by which the credibility of the predictions can be assessed. Focusing on propagation of multi-group cross-sections, the major challenge lies in the enormous size of the uncertainty space. Earlier work has explored the use of the physics-guided coverage mapping (PCM) methodology to assess the quality of the assumptions typically employed to reduce the size of the uncertainty space. A reduced order modeling (ROM) approach has been further developed to identify the active degrees of freedom (DOFs) of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. In the current work, a sensitivity study, based on the PCM and ROM results, is applied to identify a suitable compressed representation of the uncertainty space to render feasible the quantification and prioritization of the various sources of uncertainties. While the proposed developments are general to any reactor physics computational sequence, the proposed approach is customized to the TRITON-NESTLE computational sequence, simulating the BWR lattice model and the core model, which will serve as a demonstrative tool for the implementation of the algorithms.

Commentary by Dr. Valentin Fuster
2018;():V003T02A053. doi:10.1115/ICONE26-82387.

According to the development of the concept of “zero failure” or “zero fuel element defect”, accepted in 2011, which consists in reducing the number of fuel elements that are depressurized in the process of operation to the reached level in the leading countries in nuclear energy (10−6–10−5 defective fuel rods) and avoidance of fuel assemblies with non-hermetic cladding of fuel rods for further operation, including defects with a “gas leak” type, new promising fuels are being developed and introduced, including methods for justifying their safety. Thus, to ensure reliability and safety of new fuel types, it is necessary to provide procedures for monitoring current performance characteristics at all stages of the life cycle of fuel rods. In this paper, experience is given on the development and implementation of instrumentation and methods for monitoring of fuel rods with advanced types of nuclear fuel for VVER reactors that ensure the reliability, safety and competitiveness of technologies associated with the use of advanced fuel rod types, and the implementation of associated components, systems and equipment for monitoring and diagnostics. The features of the applied techniques are presented, and the new system of requirements for the implemented equipment created on their basis. This research continues, and the analysis of intermediate experimental data is carried out in this article.

Commentary by Dr. Valentin Fuster
2018;():V003T02A054. doi:10.1115/ICONE26-82397.

The main purpose of this work is to realize the internal coupling mode between the Monte Carlo neutronics transport code RMC and sub-channel thermal-hydraulic (TH) code CTF. The coupling was implemented with the coupling interfaces of RMC and CTF. And it features to use memory rather than files to transfer data from each other, which has a lot of advantages. With the internal coupling mode, power distribution calculated from RMC can be precisely provided to CTF, instead of utilizing the approximate method adopted by the external coupling mode. In addition, using memory to transfer data between those two codes can reduce the total calculation time significantly. The percentage of time reduction can be as large as 10%. Moreover, we have realized the parallel execution of CTF in the internal coupling mode and this saves a lot of time during the TH calculation.

A modified Virtual Environment for Reactor Application (VERA) Problem #6 assembly and a 4-assembly structure have been used to test the accuracy and efficiency of internal calculation mode to external calculation mode. The results show that internal coupling can give a very close solution to the external one but with 10% time reduce, and can come to an ideal convergence state with only a few iterations.

Commentary by Dr. Valentin Fuster
2018;():V003T02A055. doi:10.1115/ICONE26-82437.

Monolithic fuel is a fuel form that is considered for the conversion of high performance research reactors. In order to qualify this new fuel system, the fuel plates should meet the safety standards and perform well in reactor. The fuel system must maintain its mechanical integrity, sustain a geometric stability and should have stable and predictable irradiation behavior. The requirement to maintain mechanical integrity under normal operating conditions is primarily demonstrated by a successful testing of fuel plates up to the limiting conditions defined by the fuel performance envelope, including an adequate margin. Although large number of plates have been tested with satisfactory thermo-mechanical performance, post-irradiation examination of plates from previous RERTR-12 experiments have revealed that pillowing occurred in several plates, rendering performance of these plates unacceptable. To address such failures, efforts are underway to define the mechanisms responsible for the in-reactor pillowing, and suggest improvements to the fuel plate design and operational envelope. For this purpose, selected plates from previous experiments were simulated to understand the thermo-mechanical response of the plates to the fission density and thermally induced stresses. Simulation results were then comparatively evaluated with post-irradiation examinations of selected plates. The simulation results and experimental observations established a possible correlation between failure by plate pillowing, high porosity and a presence of tensile stress state. The study has implied that porosity leading to degradation of material properties, accompanied by a sufficiently large tensile stress state can lead to a pillowing-type failure at reactor shutdown. This paper presents these findings, discusses such failure modes, and the influence of fuel burn-up and power on the magnitude of the shutdown-induced tensile stresses.

Commentary by Dr. Valentin Fuster
2018;():V003T02A056. doi:10.1115/ICONE26-82475.

Longevity of sensors and portable devices are severely limited by temperature, chemical instability, and electrolyte leakage issues associated with conventional electrochemical batteries. Batteries undergo self-discharge and permanent loss of capacity at high temperatures, exhibit lower voltage and capacity at low temperatures, and leak electrolyte shortening operating lives, corrosion of nearby electronics, and potential safety hazards in the form of burns and poisoning. Instabilities in lithium types often short resulting explosions, fire, and venting of hazardous gasses. Betavoltaics do not have these problems and can operate in a wide temperature range without permanent degradation and loss of capacity, will not explode, and be made safe. Betavoltaic technology is maturing with advances in radioisotope sources, semiconductor materials, and developments in low-power applications, and has been identified by Department of Defense as a disruptive technology that is needed and should be pursued. This study presents experimental and modeling research on the leading betavoltaic technology, recent developments and proposed advancements. The next generation 100 microwatt betavoltaic is introduced along with projected voltage, current, and power characteristics.

Topics: Modeling
Commentary by Dr. Valentin Fuster
2018;():V003T02A057. doi:10.1115/ICONE26-82536.

In order to improve nuclear reactor’s performance and safety, a new-type fuel, Inert Matrix Dispersion Pellet (IMDP) with greatly enhanced thermal conductivity was studied. In this paper, the pellet was developed by Spark Plasma Sintering Technique (SPS), which brings an improved thermal conductivity compared with normal method and exhibits fully dense SiC matrix with a higher TRISO volume in kernel of pellet. Thermal-physical properties of the IMDP from room temperature to 1400 °C were investigated. The thermal conductivity improved 1100% at room temperature and 974% under 1000°C when TRISO volume is 40%, which behave much better than traditional UO2 fuel. The thermal diffusivity reduced and heat capacity increased at different TRISO volumes, linearly implied stable geometry at high temperature. From CT test we can see an intact structure of uniform TRISO granules. For comparison, modeling thermal conductivity is analyzed by FEM, which shows lower than measured data, indicating an optimized technology of this new method. Advantages of SPS will be discussed as well.

Commentary by Dr. Valentin Fuster

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