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Reactor Physics and Transport Theory

2010;():1-8. doi:10.1115/ICONE18-29070.

EDF R&D is developing a new calculation scheme based on the transport-Simplified Pn (SPn ) approach. The lattice code used is the deterministic code APOLLO2, developed at CEA with the support of EDF and AREVA-NP. The core code is the code COCAGNE, developed at EDF R&D. The latter can take advantage of a microscopic depletion solver expected to improve the treatment of spectral history effects. However, the direct use of the microscopic depletion solver within COCAGNE is computationally very intensive because very small evolution steps (typically 100 MWd/t) are needed to reach a good accuracy, which is not always compatible with industrial applications. In order to reduce the calculation time associated with the use of the microscopic depletion solver, a predictor-corrector scheme has been implemented within COCAGNE. It enables the use of larger evolution steps, up to 1000 MWd/t. Two kinds of tests were performed to validate this predictor-corrector scheme. Firstly, direct comparisons with APOLLO2 results were made on COCAGNE depletion calculations of fuel assemblies. These calculations involved depleting the fuel from 0 GWd/t to 60 GWd/t while keeping fixed thermal-hydraulical conditions and boron concentration, so as to be consistent with the calculation conditions of APOLLO2. It is shown that isotopic concentrations and the multiplication factor (keff) obtained by COCAGNE are consistent with APOLLO2 results. This test was also used to perform a numerical analysis of the convergence order of the new scheme. Secondly, results of the predictor-corrector scheme were compared with those of the existing calculation procedure of COCAGNE on a realistic core calculation case. The impact on global indicators such as boron concentration and cycle length was also considered. This second test indicates that, for industrial applications, the predictor-corrector procedure also gives fairly accurate results while significantly reducing the calculation time related to microscopic depletion.

Commentary by Dr. Valentin Fuster
2010;():9-15. doi:10.1115/ICONE18-29154.

Verification and validation of AEGIS/SCOPE2, a state-of-the-art in-core fuel management system for PWRs, was conducted. Verification of implementations for processes of tabulation and reconstruction of cross sections data in the system was confirmed through benchmark problems to confirm consistency between AEGIS and SCOPE2 in terms of reactivity and pin power distribution with identical computational condition in single- and multi-assembly geometry. In validation study numerical performance of the system was demonstrated in analyses of the B&W’s critical experiment and tracking calculations for commercial PWRs in combination of ENDF/B-VII.0 library. It is found the present system gives stability in prediction of critical boron concentration and radial power distribution.

Commentary by Dr. Valentin Fuster
2010;():17-25. doi:10.1115/ICONE18-29188.

The “Institut de Radioprotection et de Sûreté Nucléaire”, as the technical support of the French Safety Authority, carries out studies and research to analyze and assess the safety of all nuclear plants. In this frame IRSN studies the feasibility of modeling Material Testing Reactor core with SIMMER-III code, for simulation of reactivity initiated accidental transients. The SIMMER-III multi-physics code system was initially developed for mechanistic safety analyses of liquid metal cooled fast reactors while employing coupled spatial neutron kinetics and thermal hydraulics models. Neutronics and thermal-hydraulics SIMMER-III models have been extended to safety analyses for water cooled and moderated reactors. The use of a code like SIMMER-III requires approximations; it computes a simplified R-Z geometry and chemistry description of the core that must be validated. The methods applied consist here in developing models of the same reactor on several scales of detail. The first step is the validation of the cross section condensation for deterministic APOLLO2 calculation against Monte Carlo TRIPOLI4 2D model. Temperature effects, kinetic parameters and void coefficients on the whole core are then calculated on a 2D APOLLO2 model, using the Method of Characteristics. These parameters are also computed with a 3D combined transport and diffusion calculations by means of APOLLO2/CRONOS2 calculations, validated against a TRIPOLI4 3D precise reference model. The final step is the validation of the simmer-like R-Z geometry in APOLLO2 Sn and Pij. Finally, an R-Z geometry has been computed in SIMMER-III, for the calculation of the kinetic parameters and temperature coefficients. This validation method has been applied to Jules Horowitz Reactor, a French Material Testing Reactor currently in commissioning by the CEA. This leads to conclude that discrepancies due to simplifications are acceptable. Moreover SIMMER-III shows quite a good agreement with CEA ring calculation on the kinetic parameters. Concerning neutronics feedbacks coefficients, further analyses remain necessary.

Commentary by Dr. Valentin Fuster
2010;():27-34. doi:10.1115/ICONE18-29221.

A new lattice physics and assembly calculation code GALAXY with the 172 energy-group ENDF/B-VII.0 library has been developed. GALAXY generates few group nuclear constants used in a new core simulator COSMO-S. The GALAXY code uses the many enhanced calculation method for more explicit treatment of neutronics characteristics. The outline of enhanced methods used in GALAXY and the qualification results are shown in this paper. From the qualifications in the continuous energy Monte Carlo benchmark, critical experiment analyses and post irradiation examination (PIE) analyses, GALAXY with the library was validated and the applicability of GALAXY to PWR nuclear design was confirmed.

Commentary by Dr. Valentin Fuster
2010;():35-41. doi:10.1115/ICONE18-29252.

Monte Carlo calculation has come to be used as reference solutions instead of experiments in nuclear design code validation and verification (V&V), although comparisons with measurements are still indispensable for V&V in nuclear design. MCNP [1] is one of the most famous Monte Carlo codes widely used in the world. Many reference results are given for the analyses of critical experiments. When using the use MCNP calculations for validations of commercial design codes, we will face to a problem of lacking temperature dependent cross-sections. The cross-sections can be generated by the NJOY code [2]. However, if the model has complex temperature distribution, many NJOY calculations are necessary. Besides, if the temperature profile changes with fuel power and so on, many NJOY calculations have to be performed again and again. These back and forth procedures make us give up using MCNP for commercial LWR calculations. In order to solve this problem, we propose an easy approximation to solve the temperature problems using MCNP. Note that our technique does not require any code modifications.

Topics: Temperature
Commentary by Dr. Valentin Fuster
2010;():43-48. doi:10.1115/ICONE18-29254.

A new PWR nuclear design code system, GALAXY/COSMO-S, has been developed by Mitsubishi Heavy Industries, Ltd. (MHI) with support of Osaka University and Nagoya University. The code system has been developed based on today’s popular techniques such as MOC and semi-analytical nodal expansion method for GALAXY and COSMO-S, respectively. These codes employ several new features to improve calculation accuracy and efficiency of commercial PWR design calculations. Especially, a new robust cross section representation model based on the concept of quality engineering is introduced in GALAXY/COSMO-S system. The new representation model is intended to cover all PWR operating conditions ranging from cold shut down to hot full power. The robustness of the cross section representation model is achieved by fitting equation including cross term effect by fuel temperature, moderator density and moderator temperature. The fitting equation is determined based on Akaike’s Information Criteria (AIC) instead of empirical ways. It is the first challenge to apply the concept of quality engineering technique, which is widely used in manufacturing field, to the cross-section representation model. In order to demonstrate the applicability of COSMO-S for PWR core designs, several benchmark calculations simulating an assembly models, multi assembly models and a typical commercial plant were performed. Reactivity and power distribution were compared with the reference results prepared by GALAXY and current core design code. The results of these comparisons show excellent agreements with the references. These results validate the applicability of the GALAXY/COSMO-S system with a new cross section model for PWR core design.

Commentary by Dr. Valentin Fuster
2010;():49-54. doi:10.1115/ICONE18-29279.

The process of neutron multiplication is a discrete-time process, but the neutron transport theory takes neutron multiplication as a continuous neutron source, which ignores the discrete-time process of neutron multiplication, which would take in errors, so it is necessary for describing the process of neutron multiplication as a discrete-time process. “The neutron doubling formula including delayed neutrons” has been established which describes the process of neutron multiplication as a discrete-time process, but it has nothing to do with space. “The neutron doubling formula including delayed neutrons” could not be used to describe the variety of distributing of neutron density in transient process; it also could not be used to deal with the problem of three-dimensional space. In order to solve the problems mentioned above, the space-time neutron multiplication formula is established. Based on the theory of neutron multiplication, the concept of space is introduced to the neutron multiplication formula and the space-time neutron multiplication formula is established by taking into account of neutron transport. The formula can describe the inherent physical process of neutron multiplication in fission chain reaction system. The test of space-time neutron multiplication formula is done, which proves the formula is right. Given the initial neutron density as well as the multiplication factor, the formula can strictly describe the variety of neutron density (neutron flux density) with time. It could be used for setting a standard for estimating error for the measurement of neutron flux density as well as numerical calculation; the space-time neutron multiplication has larger applicability compared with the “neutron doubling formula including delayed neutrons”.

Topics: Neutrons , Spacetime
Commentary by Dr. Valentin Fuster
2010;():55-62. doi:10.1115/ICONE18-29285.

Most startup and shutdown operations in advanced boiling water reactors (ABWRs) are automated by an automatic power regulator (APR). Hitachi and Hitachi-GE utilized the three-dimensional transient analysis code TRACG to design and verify the APR control algorithms. To verify the algorithms, an external neutron source model that makes it possible to simulate a sub-critical initial core, a water temperature reactivity model, a startup range neutron monitor (SRNM) model, and the APR system models were developed and coded onto the TRACG code. The improved TRACG code has been tested and verified with ABWR startup test data. In the test, the criticality was achieved 40 min after beginning of control rod (CR) withdrawal. The code results, for example, CR operation timing, CR withdrawal length, and signals of the neutron sensors agreed well with the test data. In the heat-up control mode, the measured increasing rate of the reactor water temperature was well simulated for a period longer than six hours.

Topics: Heat
Commentary by Dr. Valentin Fuster
2010;():63-68. doi:10.1115/ICONE18-29315.

The core flux (power) distribution is very important to safe and economical operation of nuclear reactor. It can be obtained by many methods depending on the desired accuracy and execution time. For on-line core surveillance and regulation, we need to get the real-time flux distribution. If the true local parameters such as fuel temperature, coolant temperature and material density were known, the solution of the diffusion equation with instantaneous parameters could, in principle, provide the necessary spatial details. However, in reality, it is impossible to obtain the operational “readings” of these parameters for each fuel cell. The detector results at certain locations can be applied to improve the results of the only diffusion calculations by Flux Mapping methods. Function expansion method is employed to express the approximate real distribution by the combination of several Flux Mapping method results as the expansion basis functions. The Harmonics Synthesis Method (HSM) and Least-Square method are combined to get a new Flux Mapping method in this paper. The simulation results show that the new method can be used for Flux Mapping and get better results.

Commentary by Dr. Valentin Fuster
2010;():69-77. doi:10.1115/ICONE18-29373.

A CANDU-SCWR core is designed by using a 3D neutronics/thermal-hydraulic coupling method. In the fuel channel design, a typical 43-element fuel bundle is used, the coolant and the moderator are supercritical water and heavy water respectively. The thickness of the moderator is optimized to ensure the negative coolant coefficient during operation. With 1220 MW electric power, the reactor core is designed with a diameter of 4.8m and length of 4.95m, and there are totally 300 fuel channels, each of which consists of 10 fuel bundles. The inlet coolant temperature is set to be 350 °C °C and the operation pressure is 25 MPa. In order to flatten the radial power distribution, the loading pattern of the equilibrium cycle is optimized, and an optimized fuel management scheme is used with three batches refueling, burnable poison Dy2 O3 is used to flatten the power peaking. The numerical results show that the average power density is 42.75 W/cm3 , while the maximum linear element rate (LER) is 575W/cm. The average discharged burnup of the equilibrium is 48.3GWD/tU, and a high average outlet coolant temperature of 625 °C is achieved with a maximum cladding surface temperature less than 850 °C. Besides, the coolant temperature coefficient is negative throughout the cycle.

Commentary by Dr. Valentin Fuster
2010;():79-87. doi:10.1115/ICONE18-29420.

The Generalized Minimal RESidual (GMRES) method, which is a widely-used version of Krylov subspace methods for solving large sparse non-symmetric linear systems, is adopted to accelerate the 2D arbitrary geometry characteristics solver AutoMOC. In this technique, a formulism of linear algebraic equation system for angular flux moments and boundary fluxes is derived as an alternative to traditional characteristics sweep (i.e. inner iteration) formalism, and then the GMRES method is implemented as an efficient linear system solver. Several numerical results demonstrate that the acceleration technique based on Krylov subspace methods can be applied to arbitrary geometry MOC solver successfully, and may obtain higher efficiency than the original characteristics solver does because of its spectacular effect on reducing both the number of outer iterations and the total computing time. Moreover, the results could be improved by Lyusternik-Wagner extrapolation technique in some cases.

Topics: Geometry
Commentary by Dr. Valentin Fuster
2010;():89-96. doi:10.1115/ICONE18-29425.

Due to the need for high accuracy and more complex core geometry, whole-core transport calculations with the MOC are demanded. Recently, a 2D+1D method has been proposed to perform 3D whole-core transport calculations, but it is not easy to apply this method to the cases with complex geometry in the axial direction. Direct 3D MOC calculations are thus required. This approach is however limited by the amount of computer memory which is required for the ray tracing data that is required by the MOC. This paper presents a modular tracing technique and corresponding angular quadrature sets, which can greatly reduce the amount of ray tracing information, making it practical for the MOC to deal with the large 3D problems. In this paper, we introduce the model of the modular ray tracing technique and give some results of our method.

Commentary by Dr. Valentin Fuster
2010;():97-105. doi:10.1115/ICONE18-29429.

This paper describes a one-dimensional wavelet-based spatial discretization scheme for the first-order neutron transport equation. Two special features are introduced: i) the spatial variable is discretized using the Daubechies’ wavelets on the interval, and the neutron flux is represented in term of the wavelet series in a normalized node, the tradition SN angular discretization scheme is used in solving the equation, and ii) the wavelet Galerkin method is applied here, using the Daubechies’ scaling function as both the trialing function and weighting function, the integrations of Daubechies’ scaling function and its derivative in the Galerkin system are calculated numerically, using the difference quotient instead of the derivative. The boundary conditions and interface conditions are given in the exact form of wavelets series and added into the Galerkin system in special locations. The LU decomposition method is applied to solving the matrix in formed in the Galerkin system. The test results on several benchmark problems indicate that the wavelet-based spatial discretization scheme in this paper is capable of handling the first-order neutron transport equation, accurate in treating the boundary condition while using the wavelets expansion in spatial discretization, effective in treating the transport problems in the deep penetrating medium and in strong heterogeneous medium.

Topics: Neutrons , Wavelets
Commentary by Dr. Valentin Fuster
2010;():107-112. doi:10.1115/ICONE18-29439.

The EPR™ reactor has been designed by AREVA to support economical fuel cycles. The progress in the reactor and systems design improves the reactor safety, and allows the EPRTM reactor to support the large range of high performance fuel management strategies covering cycle length from 12 to 24 months. Different fuel management strategies with 12, 18 and 24 month cycles are described. Economic analyses are performed to illustrate the low uranium consumption and the high fuel cycle performance compared with the fuel managements implemented in most current traditional PWR reactors.

Topics: Fuels , Cycles
Commentary by Dr. Valentin Fuster
2010;():113-117. doi:10.1115/ICONE18-29442.

The spherical harmonics (Pn) finite element method, the Sn finite element method, the triangle transmission probability method and the discrete triangle nodal method were all introduced to solve the neutron transport equation for unstructured fuel assembly respectively. The computing codes of each method were encoded and numerical results were discussed and compared. It was demonstrated that these four methods can solve neutron transport equations with unstructured-meshes very effectively and correctly, they can be used to solve unstructured fuel assembly problem.

Commentary by Dr. Valentin Fuster
2010;():119-125. doi:10.1115/ICONE18-29481.

In the spent fuel dissolving vessel, the resonant isotopes in fuel pellets is surrounded by fissile solution. Complicated resonance interaction between the solid fuel and fissile solution is induced. Conventional self-shielding methods based on the equivalence theory for PWR cell and assembly will not apply because the neutron flux spectrum in the fissile solution are quite different from those PWR cases where no resonant isotope is in the moderator. In the present paper, Sub-group method is applied to calculate the space-dependent self-shielded multi-group cross-section. Sub-group parameters such as sub-group weights and sub-group cross-sections are calculated by fitting method using the NJOY processed multi-group constant under various dilutions. Elastic scattering Resonance for heavy isotopes is taken into account. Space-dependent sub-group flux is then calculated by fix source sub-group transport calculation. Spatially-dependent self-shielded multi-group cross-sections and multiplication factor can be obtained. Calculations are carried out on the NEA-CRP proposed standard problems for fissile pellets in fissile solution. Results show that the sub-group method is a very promising deterministic method for the self-shielding calculation in the spent fuel dissolution.

Commentary by Dr. Valentin Fuster
2010;():127-131. doi:10.1115/ICONE18-29482.

This paper describes a newly developed Monte Carlo code used for reactor analysis called RMC1.0, which is based on ACE format library. RMC1.0 is able to estimate criticality eigenvalue, and tally flux/spectrum with collision estimation method or tracking length method. Series of benchmarks and other examples are calculated for validation, which prove that RMC1.0 gives a good performance in both accuracy and efficiency compared with mcnp5. Despite its limitation in geometry processing, RMC1.0 has made a profitable attempt in self-development of Monte Carlo code for reactor analysis.

Commentary by Dr. Valentin Fuster
2010;():133-141. doi:10.1115/ICONE18-29491.

Resonance self-shielding calculation is very important in reactor physics calculation. Conventional resonance calculation method has some fundamental defects, which hinders its application in some problems. The Hyperfine Energy Group Resonance Calculation Method is studied in this paper and a code named UFOP is developed based on this method. In this method, the resonance energy range is divided into hyperfine energy intervals (tens of thousands) and the collision probabilities are calculated. Then the slowing-down equation is directly solved based on CPM (collision probability method). Some techniques are applied in solving the slowing-down equation for improving computational efficiency and reducing calculation error. A resonance benchmark problem with homogeneous and infinite material is calculated to validate the accuracy of the computation code and the hyper-fine group cross-section library utilized in the code. A PWR fuel cell is also calculated and the results are compared with MCNP. The results show good accuracy of this method and the validity of UFOP code.

Topics: Resonance , Neutrons
Commentary by Dr. Valentin Fuster
2010;():143-150. doi:10.1115/ICONE18-29498.

The radial power distribution within the fuel rod is important in fuel integrity evaluation. In this paper, the wavelets scaling function expansion method is applied to evaluate the radial power distribution and obtain continuous-energy spectrums in the fuel rod with a temperature distribution. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are processed by NJOY, while the multi-group nuclear data library is jeff31 issued by IAEA. However, for different temperature problems, because of Doppler effect, the continuous-energy data library is different, and should be updated by applying NJOY. The calculation efficiency is a problem needed to be improved. Therefore, in this paper, the interpolation is utilized to obtain the continuous-energy cross-sections for other temperatures between those given temperatures. Also the precision of the temperature interpolation is discussed. Finally, the differences of continuous-energy spectrums, reaction rates and k-inf results are presented and compared with MCNP calculation.

Commentary by Dr. Valentin Fuster
2010;():151-157. doi:10.1115/ICONE18-29518.

An analytic basis function expansion nodal method for directly solving the two-group neutron diffusion equation in the triangular geometry is proposed in the present paper. In this method, the distribution of neutron flux is expanded by a set of analytic basis functions. The diffusion equation is satisfied at any point in a triangular node for each group assuming that the flux within a node is flat. No transverse integration is needed. To improve the nodal coupling relations and computation accuracy, nodes are coupled with each other fulfilling both the zero- and first-order partial neutron current moments across all the three interface of the triangle mesh at the same time. Coordinate conversion is used to transform arbitrary triangle into regular triangle in order to simplify the derivation. A new sweeping scheme is developed for the triangular mesh and the response matrix technique was used to solve the nodal diffusion equation iteratively. Based on the proposed model, the code ABFEM-T is developed. Validation of code for accuracy and efficiency are carried out by calculating both rectangular and hexagonal assembly benchmark problems. Numerical results for the series of benchmark problems show that both the multiplication factor and nodal power distribution are predicted accurately. Therefore this method can be used for solving neutron diffusion problems in complex unstructured geometry.

Commentary by Dr. Valentin Fuster
2010;():159-165. doi:10.1115/ICONE18-29533.

Monte Carlo codes are powerful and accurate tools for reactor core calculation. Most Monte Carlo codes use the point-wise data format, in which the data are given as tables of energy-cross section pairs. When calculating the cross sections at an incident energy value, it should be determined which grid interval the energy falls in. This procedure is repeated so frequently in Monte Carlo codes that its contribution in the overall calculation time can become quite significant. In this paper, the time distribution of Monte Carlo method is analyzed to illustrate the time consuming of cross section calculation. By investigation on searching and calculating cross section data in Monte Carlo code, a new search algorithm called hash table is elaborately designed to substitute the traditional binary search method in locating the energy grid interval. The results indicate that in the criticality calculation, hash table can save 5%∼17% CPU time, depending on the number of nuclides in the material, as well as complexity of geometry for particles tracking.

Commentary by Dr. Valentin Fuster
2010;():167-174. doi:10.1115/ICONE18-29540.

Using NJOY to generate the temperature dependent neutron cross-section is too time-consuming in practice, especially for many nuclides. So an approach involving interpolation between nuclear data libraries at different temperatures is investigated. Based on the ACE data at different temperatures, we used ITND — an neutron cross-section interpolation program, to generate the target temperature ACE data, then we compared it with the ACE data which generated by NJOY at the same temperature. We focused on the interpolation result of 238U, 235U, 232Th, Zr, 16O, 10B and 1H at the temperature of 575K. To that nuclides, several interpolate schemes were studied, and we demonstrated the relative differences, and explain their reasons. Finally we applied these ACE data to benchmark calculation, and good agreement was observed with the benchmark results.

Commentary by Dr. Valentin Fuster
2010;():175-180. doi:10.1115/ICONE18-29547.

Metal fuelled sodium cooled fast reactors are known to have high breeding ratio and short doubling time. Due to these features they play a very important role in the energy scenario, where higher power growth is required. Large sodium cooled fast reactors have positive sodium void coefficient, which is considered to be undesirable feature even though reactor safety can be established for all design based accidents like loss of flow and transient over power accidents. These types of fast reactors, which have harder neutron spectra are having higher sodium void coefficient compared to ceramic fuelled fast reactors. In many of the safety analysis the total sodium void is calculated and it is used in the safely evaluation. However the sodium in the metal fuelled reactor has got three parts, namely bonding sodium, coolant sodium and the sodium in the inter space of subassembly hexagonal cans. In the reactor accident scenario the behavior of these three components of sodium will be different and will effect the sequence of the accident. The finer details, of the fuel sub assembly, are modeled in to Monte Carlo code and the sodium void coefficient is calculated for each of the component for the fuel zones. This study will be helpful in improving safety of the reactor and also reducing the conservatism in the safely features.

Topics: Sodium
Commentary by Dr. Valentin Fuster
2010;():181-188. doi:10.1115/ICONE18-29548.

Nuclear data library is the cornerstone in the nuclear reactor’s design and calculation. The WIMS-D multi-group library and ACE format library (mainly used in MCNP) is applied frequently in the nuclear calculation. We have developed a new self-shielding calculation procedure based on Wavelets scaling function expansion method. This procedure needs several parts in both WIMS-D and ACE format library. So the consistency of two libraries becomes a very serious problem. This may bring in large errors. In this paper, NJOY cross section processing system is used to produce new WIMS-D and ACE format library from the same ENDF/B data. We compute some homogenous problems using new and old libraries in WIMS-D and ACE format. The results of the two new libraries and the old libraries are compared respectively. It is found that there are consistency problems between the two libraries. The newly produced libraries are more compatible than the old ones.

Topics: Design , Errors , Wavelets
Commentary by Dr. Valentin Fuster
2010;():189-196. doi:10.1115/ICONE18-29562.

In this paper, localization of a noise source from limited neutron detectors sparsely distributed throughout the core of a typical VVER-1000 reactor is investigated. For this purpose, developing a 2-D neutron noise simulator for hexagonal geometries based on the 2-group diffusion approximation, the reactor dynamic transfer function is calculated. The box-scheme finite difference method is first developed for hexagonal geometries, to be used for spatial discretisation of both 2-D 2-group static and noise diffusion equations. The dynamic state is assumed in the frequency domain which leads to discarding of the time disrcetisation. The developed 2-D 2-group neutron noise simulator calculates both the discretised forward and the adjoint reactor transfer function between a point-like source and its induced neutron noise, by assuming the noise source as an absorber of variable strength type. Benchmarking of the mentioned neutron noise simulator revealed that it works satisfactorily. Finally, by using the inversion method of reconstruction, the location and values of a noise source of the type absorber of variable strength (or reactor oscillator) in VVER-1000 reactor cores are determined. Accuracy of this method is highly acceptable.

Commentary by Dr. Valentin Fuster
2010;():197-204. doi:10.1115/ICONE18-29563.

Transmutation of the nuclear waste which contains MA and FP is important and necessary. Because the major MA and FP’s nucleus densities is large in transmutation calculation, now existing multi-group WIMS libraries are not suitable for transmutation calculation. This paper makes a study for the parameters value choice such as weight spectrum and background cross-section σ0 in NJOY. With the new parameters value, produce suitable multi-group data of FP and a new library for transmutation calculation. In purpose of comparing these two libraries, build a simple transmutation cell model in HFETR. Achieve the cell calculation separately by WIMSD with new library and original library. Then compare the results of parameters such as K-INF, flux, and absorption reactions. The results show that the values of these parameters are different in resonance energy range, so it is necessary to think over the resonance of FP when a multi-group library is produced for thermal reactor transmutation.

Commentary by Dr. Valentin Fuster
2010;():205-210. doi:10.1115/ICONE18-29614.

A high performance hardware acceleration coprocessor built on field programmable arrays (FPGAs) is designed to accelerate neutron transport computation for three dimensional whole reactor cores. The acceleration coprocessor is designed based on the reconfigurable computation techniques and adopts the dataflow-driven non von Neumann architecture for high efficient parallel computation. The hardware acceleration coprocessor supports much more intensive available computation power compare with the same-era CPUs, and is compatible with existing software acceleration methods. It reaches about 20 times speed up in simulation validations. It is the first time that the reconfigurable hardware acceleration techniques are used to improve the computational efficiency of the reactor physics and neutron transport simulations.

Commentary by Dr. Valentin Fuster
2010;():211-216. doi:10.1115/ICONE18-29638.

A new approach based on the method of characteristics (MOC) and Rosenbrock method is developed to solve the time-dependent transport equation in one-dimensional (1D) geometry without any approximation and considering delayed neutrons. Within the MOC methodology, the leakage term in time-dependent transport equation can be simplified to spatial derivative of the angular flux along the characteristics lines. For 1D geometry, the proposed exponential correlation derived from the steady-state MOC equations provides the correlation between the cell outgoing angular flux and the cell average angular flux. Thus, the spatial derivative term can be further substituted by the relation containing only the cell average angular flux that represents the unknowns. Therefore, the 1D time-dependent transport equation is decomposed into a series of locally coupled ordinary differential equations (ODE). Rosenbrock method was chosen to solve the system of ODEs. It is a fourth order explicit method with automatic step size control feature developed for stiff ODEs. The FORTRAN90 numerical program is developed to thus solve the time-dependent transport equation considering delayed neutrons in 1D geometry with both vacuum and reflective boundary conditions. The step perturbation is currently supported. The method presented in this paper was verified in comparison to 1D fast reactor benchmark showing good accuracy and efficiency.

Topics: Equations
Commentary by Dr. Valentin Fuster
2010;():217-226. doi:10.1115/ICONE18-29654.

The decay heat (energy due to decay of unstable nuclei) is a small fraction of reactor power at nominal conditions, but after reactor shut-down it is the most important heat source. For taking this source into account in design and safety studies, recommendations are available for fuels of operating reactors, such as UOX and MOX. Fuels for EFIT (European Facility for Industrial Transmutation), unlike UOX and MOX, should contain a significant amount of Minor Actinides (MAs) that would influence decay heat. CEA, CIEMAT, ENEA, FZK (now KIT), PSI and SCK•CEN established a benchmark case and computed decay heat curves for MA-bearing fuels and a MOX-type fuel. The decay heat in the fuels with MAs is appreciably higher than in MOX, except for low burnup cases after short cooling times. This should be taken into account in the design of the decay heat removal system for EFIT. The obtained differences between the decay heat in MA-bearing and MOX fuels are supposed to be representative for the benchmark (or similar) conditions. More effort is needed to evaluate the uncertainties of the computed results.

Topics: Heat , Fuels , Uranium
Commentary by Dr. Valentin Fuster
2010;():227-236. doi:10.1115/ICONE18-29691.

In this paper, reconstructed local fission rates obtained with the two-group nodal diffusion program PRESTO-2, used at the Leibstadt Nuclear Power Plant (KKL) in Switzerland, are compared with experimental results and MCNPX calculations. The experimental facility consists of a test zone, where the measurements are made, surrounded by a buffer zone and two driver zones that render the system critical and also contain the control rods. The test zone consists of a tank that contains a 3×3 array of BWR fuel assemblies of type SVEA-96+. Four cases are considered, all corresponding to a 1.23 m high active zone moderated with light water at room temperature: 1) axially uniform enrichment and gadolinium content, 2) like case 1 but with a L-shaped control blade completely inserted, 3) enrichment and gadolinium content change at the core mid-plane, 4) like case 2 but with the control blade partially inserted. The comparisons give insight into the accuracy of the pin power reconstruction methodology. The axially uniform case without control blade shows a good radial agreement and a well predicted axial curvature of the flux. On the other hand, systematic deviations are observed in the radial direction for the controlled cases, with the axial heterogeneities causing deviations around the discontinuity and also in the axial curvature of the flux.

Commentary by Dr. Valentin Fuster
2010;():237-243. doi:10.1115/ICONE18-29709.

A method of on-line monitoring for commercial PWRs using eigenfunctions has been proposed in prevenient works. In this method, the eigenfunctions combine with the detector readings are used to reconstruct the real reactor power distribution. But it is very difficult to choose the eigenfunctions because the condition of the reactor is much complex. Therefore, some improvements on this method are studied in this paper. A number of representational conditions according to the reactor fuel management are picked up to create a data library of different eigenfunctions. On the monitoring process, the computer will judge that which represen6tational conditions most close to the real reactor condition and choose the eigenfunctions which will be used to reconstruct the reactor core power distribution combine with the detector readings. A reactor of Qinshan Nuclear Power Corporation is studied here as an example. The detector readings are from the simulator. The numerical result shows that this method can reconstruct the reactor power distribution with high speed (about 0.03 seconds for each step) and high accuracy (the relative errors are lower than 3% mostly).

Topics: Eigenfunctions
Commentary by Dr. Valentin Fuster
2010;():245-250. doi:10.1115/ICONE18-29763.

AECL has developed enhanced versions of the reactor physics computer codes for analysis of CANDU® reactors and the ACR-1000™. The central codes that comprise the analysis toolset are WIMS-AECL (a lattice code), RFSP (a core code) and MCNP5 (a Monte Carlo code). The toolset, with ENDF/B-VI nuclear data, has been validated for application to the ACR-1000 design. In addition to comparisons of code predictions against relevant experiments conducted in AECL’s ZED-2 critical facility, advanced methods based on cross-section sensitivity/uncertainty (S/U) analysis were used to extend the results of bias and uncertainty in reactivity coefficients, derived from analysis of ZED-2 tests, to the ACR-1000 reactor. The validation of this toolset with ENDF/B-VII nuclear data is proposed for application to analysis of a Thorium-fuelled CANDU Reactor (TCR). The TCR is based on the Enhanced CANDU 6™ (EC6™) reactor [1] and would operate with a fuel design that incorporates both low-enriched uranium (LEU) oxide and thorium oxide fuel elements in the same fuel bundle to achieve enhanced fuel and core performance with thorium fuel. For the initial TCR toolset qualification, important reactor physics phenomena would be validated using several relevant ZED-2 experiments performed in the past. Results from experiments with a variety of oxide fuels are available, including plutonium/thorium (Pu/Th), 233 U/Th, 235 U/Th, LEU and CANDU-MOX (containing a mixture of plutonium, uranium and dysprosium to simulate the reactor physics affects of fuel burnup). Taken together along with other relevant experimental data, these experiments would be expected to address the important isotopes and many phenomena for the TCR and to enable the validation of the reactor physics toolset for this design. Additional confirmatory experiments would reduce uncertainties. This paper describes the qualification process, including validation, which is proposed to support the use of the reactor physics toolset for the TCR.

Topics: Physics
Commentary by Dr. Valentin Fuster
2010;():251-258. doi:10.1115/ICONE18-29971.

The Monte Carlo (MC) simulation method, known to handle complex problems which may be formidable for deterministic methods, will always require validation with classic problems that have evolved historically from deterministic methods [1–5] based on Chandrasekhar’s method in radiative transfer, Fourier transforms, Green’s functions, Weiner-Hopf method etc which are restricted to simple geometries, such as infinite or semiinfinite media, and simple scattering laws too for practical application. This work compares deterministic results with MC simulation results for neutron flux in a slab. We consider mono-energetic transport problem in an infinite medium and in a 1-D finite slab with isotropic scattering. The transport theory solutions used in infinite geometry are the Green’s function solution and the spherical harmonics (P1 , P3 ) solutions, while for the 1-D finite slab, we refer to a transport benchmark for which an exact solution is available. For diffusion theory, we consider the Green’s function infinite geometry solution, and the exact and eigen-function numerical solution for finite geometry (1-D slab). The objective of this work is to illustrate the results from all the methods considered especially near the source and boundaries, and as a function of the scattering probability. The results are plotted for six elements that include a strong absorber, such as gadolinium, and a strong “scaterrer” such as aluminium. The present work is didactic and focuses on problems which are simple enough, yet important, to illustrate the conceptual difference and computational complexity of the deterministic and stochastic approaches.

Commentary by Dr. Valentin Fuster
2010;():259-265. doi:10.1115/ICONE18-30017.

This paper is to evaluate the depletion capability of the DeCART code for PMR200 two-dimensional core. DeCART solves the burnup equation by the matrix exponential based on the Krylov Subspace method. The depletion calculation is performed up to 270 EFPD by using total 14 burnup steps. The double heterogeneity effect is resolved by introducing the RPT method. The DeCART solutions by using a 47-G PWR neutron library are compared with the McCARD Monte Carlo solution which uses ENDF/B-VII. DeCART shows about maximum 1200 pcm eigenvalue, about maximum 2.5% block and 6.0% pin power differences near 180 EFPDs. The differences are mainly due to the library and the geometrical model discrepancy. While the reference calculation is performed by imposing the vacuum condition for the vessel outside, DeCART uses a zigzag-type boundary model. The use of the VHTR neutron library that is scheduled to develop reduces the eigenvalue differences. In the computing time, DeCART requires about 2 and half hours for 14 depletion steps. Therefore, it is concluded that the DeCART code produces a reasonable result for the VHTR core depletion calculation within an affordable computing time.

Commentary by Dr. Valentin Fuster
2010;():267-273. doi:10.1115/ICONE18-30046.

Applicability and efficiency of Tone’s method used for resonance calculations is improved by incorporating the Method of Characteristics (MOC). Verification calculations are carried out in three configurations, i.e., infinite slab, infinite cylinder, and pin-cell. The validity of the modified Tone’s method has been confirmed through the results of verification calculations. Since MOC has excellent geometric flexibility and efficiency for large and complicated configurations, the proposed method can be applied to a large, complicated, and general geometry such as a whole fuel assembly or a full core configuration.

Topics: Resonance , Geometry
Commentary by Dr. Valentin Fuster
2010;():275-280. doi:10.1115/ICONE18-30159.

This paper presents a new approach for fuel burn up evaluation and radioactive inventory calculation used in Tehran Research Reactor. The approach is essentially based upon the utilization of a program written by C# which integrates the cell and core calculation codes, i.e., WIMSD-4 and CITVAP, respectively. Calculation of fuel burn up and radioactive inventories has been done for 26 core configuration of Tehran Research Reactor with HEU fuel element. The present inventory and fuel enrichment of each fuel element have been calculated.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2010;():281-286. doi:10.1115/ICONE18-30160.

In this research, new software package for neutronic calculations, especially kinetic parameters of PWR reactors, has been developed. The program used to link the WIMS-D5, BORGES and CITVAP nuclear codes has been written in Visual C# programming language. This software was used for calculation of kinetic parameters of WER-1000 and NOK Beznau reactors. The ratios (βeff )i /(βeff )core of parameters, which are an important input data for the reactivity accident analysis, were also calculated. The results were compared with final safety analysis report (FSAR) and published documents.

Commentary by Dr. Valentin Fuster
2010;():287-295. doi:10.1115/ICONE18-30195.

In today’s cross section data processing process, asymptotic scattering model is employed by NJOY for the neutron/nucleus elastic scattering interactions in the epithermal energy region, which means that the energy of a scattered neutron is always lower than its incident energy and it falls evenly within the interval of [αE, E]. This model has recently been proved to have non-ignorable errors at some resonances of heavy nuclides. In this study, to investigate the impact of heavy nuclides resonance elastic scattering models to the resonance integrals, exact scattering kernel is employed and a deterministic code Estuary is developed to efficiently solve the neutron slowing down problem. Numerical results demonstrate that with the use of Estuary, results given in the literature obtained by the Monte Carlo method can be reproduced. With the resonance cross section approximately represented by the single-level Breit-Wigner formulation, investigations are made for different resonance parameters for both asymptotic and exact scattering models. Relations between errors and these related parameters are summarized.

Commentary by Dr. Valentin Fuster
2010;():297-301. doi:10.1115/ICONE18-30203.

In TRAC-BF1 the cross-sections are specified in the input deck in a polynomial form. Therefore, it is necessary to obtain the coefficients of this polynomial expansion. One of the methods proposed in the literature is the KINPAR methodology developed in the UPV. This methodology uses the results from different perturbations of the original state to obtain the coefficients of the polynomial expansion. The simulations are performed using the SIMULATE3 code. In this work, a new methodology called SIMTAB-1D to obtain the cross-sections sets in 1D is presented. The first step consists of the application of the SIMTAB methodology, developed in the UPV, to obtain the 3D cross-sections sets from CASMO4/SIMULATE3. These 3D cross-sections sets are collapsed to 1D, using as a weighting factor, the 3D thermal and rapid neutron fluxes obtained from SIMULATE3. This new methodology will be applied to the simulation of the turbine trip transient in Peach Bottom NPP using the TRAC-BF1 code. The results of the steady state in TRAC-BF1 using the KINPA R methodology and the new methodology are compared with the reference SIMULATE3 results.

Commentary by Dr. Valentin Fuster
2010;():303-309. doi:10.1115/ICONE18-30212.

The ANSI/ANS-6.4.3-1991 Standard, Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials, is currently being updated by an American Nuclear Society (ANS) Working Group. The ANSI/ANS-6.4.3, 1991 standard, which is “withdrawn” due to the failure to meet the requirement of the ANS to have standards updated every ten years, contains buildup factor values that are derived from data that is over seventeen years old. In addition, computer technology has significantly improved since 1991, allowing for more complicated, computationally demanding codes to be utilized. Therefore, the ANSI/ANS-6.4.3-1991 standard is being re-visited to include updated data and to include modern codes. Gamma-ray buildup factors and attenuation coefficients are being generated for common shielding materials (e.g., concrete, steel, water, etc.) utilizing the ENDF/B-VI.8 cross-section data library, which is distributed by Brookhaven National Laboratory’s National Nuclear Data Center (NNDC). One modern code, Monte Carlo N-Particle (MCNP5/MCNPX), is being compared with PALLAS-1D (VII) and Anisotropic Source-Flux Iteration Technique (ASFIT-VARI), which were used to develop values included in ANSI/ANS-6.4.3-1991. PALLAS-1D (VII) is a code for direct integration of transport equation in one-dimensional plane and spherical geometries while ASFIT-VARI is a gamma-ray transport code system for one-dimensional finite systems. MCNP5 is a general purpose Monte Carlo radiation transport code that tracks particles (e.g., neutron and photons) at numerous energies in a three dimensional configuration of materials. The MCNP5 radiation transport code require response function input to provide dose and exposure output. Mass energy-absorption coefficients and mass energy-transfer coefficients are required to develop absorbed dose and exposure responses. The National Institute of Standards and Technology (NIST) provide and maintain the most up-to-date mass energy-absorption coefficient and mass energy-transfer coefficient database currently available. The NIST values are used in these initial buildup factor calculations to prove the validity of the methodology used and allow for preliminary comparisons. The energy absorption (dose in material) and exposure buildup factors are calculated at mfp values of interest, consistent with ANSI/ANS-6.4.3-1991 up to forty mfp. Comparisons between the new buildup factors and the previous results presented in the ANSI/ANS-6.4.3-1991 standard indicate that there is fairly good agreement. Differences in buildup factor values can be attributed to differences in cross-section data libraries, numerical methods, and physics treatments within the respective codes.

Commentary by Dr. Valentin Fuster
2010;():311-316. doi:10.1115/ICONE18-30310.

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

Commentary by Dr. Valentin Fuster
2010;():317-322. doi:10.1115/ICONE18-30370.

The continuous-energy resonance self-shielding calculation method based on wavelets scaling function expansion method is a valuable potential method to solve the complex resonance problem. Within the fast and thermal energy ranges, the standard multi-group treatment is applied, while Daubechies’ wavelets scaling function expansion method is used to discretize the energy variable of neutron flux within the resonant energy range. In this method, the neutron transport equation is transformed to a set of expansion coefficients equations of wavelets scaling functions. Calculation of the coefficients is very time consuming so that a powerful neutron transport calculation method is needed for better calculation efficiency. In this paper, the discrete direction probability method (DDPM) is employed as a tool for solving the wavelet scaling function expansion coefficients. The DDPM combines the desirable features of interface current method as well as the method of characteristics.

Topics: Resonance , Wavelets
Commentary by Dr. Valentin Fuster

Nuclear Education, Public Acceptance and Related Issues

2010;():323-330. doi:10.1115/ICONE18-29183.

Since September 2008, a new Master’s programme in Nuclear Engineering (NE) is being offered jointly by the two Swiss Federal Institutes of Technology, viz. EPFL at Lausanne and ETHZ at Zurich. The present paper discusses the development and running of one of the key compulsory courses, viz. Reactor Experiments. This is centred mainly around the utilization of EPFL’s teaching reactor, CROCUS. The course is of seven weeks’ duration, with 10 contact hours per week, and is held during the second half of the first semester. Following an introductory session, the class is divided into six groups. With a total of 12 experiments to be carried out by each group, there are usually 2 half-day experiments to be conducted by a given group per week. Effectively, six experiments have to be running simultaneously at any given time. There are two basic types of experiments which constitute the course contents, viz. those related to radiation measurements and neutronics phenomena, and those conducted using the reactor itself. Details are provided concerning the practical realisation of the various experiments. For example, certain difficulties encountered in the context of achieving adequate counting statistics are discussed, and the corresponding solutions presented. The course has been conducted twice to date, and the feedback from the students has been very positive.

Commentary by Dr. Valentin Fuster
2010;():331-340. doi:10.1115/ICONE18-29218.

The two national technical universities in Switzerland, viz. the Swiss Federal Institutes of Technology at Lausanne (EPFL) and at Zurich (ETHZ) have a rich and long tradition in nuclear education. Student research in nuclear engineering, particularly at the doctoral level, has usually been conducted in collaboration with the Paul Scherrer Institute (PSI) at Villigen, the national research centre where most of the country’s fission-related R&D is carried out. A significant part of this R&D is carried out in close collaboration with the Swiss Nuclear Utilities (swissnuclear). The four above, key national players in nuclear teaching and research in Switzerland — EPFL, ETHZ, PSI and swissnuclear — have recently pooled resources in implementing a new Master of Science degree in Nuclear Engineering (NE). The present paper describes the main features and experience acquired to date in the running of this, first-ever, common degree offered jointly by the two Swiss Federal Institutes of Technology. The program, although naturally addressing Switzerland’s needs, is clearly to be viewed in an international context, e.g. that of the Bologna Agreement. This is reflected in the composition of the first two batches, with about 70% of the students having obtained their Bachelor degrees from universities outside Switzerland. Starting September 2010, the curriculum of the EPFL-ETHZ NE Master will be upgraded, from its current 90 ECTS credit points (3 semesters) to a 120 ECTS (4 semesters) program. An overview is provided of the current 90-ECTS curriculum, as also a sketch of the changes foreseen in going to 120 ECTS.

Commentary by Dr. Valentin Fuster
2010;():341-346. doi:10.1115/ICONE18-29271.

Public participation to the decision of nuclear issues has been progressing in the already developed countries of nuclear power such as USA, West Europe and Japan. Risk communication by the nuclear industries to the society is time consuming and high cost. The authors of this paper have proposed the frame of two way risk communication for the nuclear industries to cope with the difficulty of social acceptance. In this paper, the author investigated the case of the largest nuclear power station in Japan which had hit by large earthquake in July 16, 2007, being back to the normal operation, with continuous plant restoration work and hard efforts to get approval from local government of commercial operation, from this frame of two way risk communication.

Commentary by Dr. Valentin Fuster
2010;():347-353. doi:10.1115/ICONE18-29327.

Perception that U.S. government energy subsidies have favored nuclear energy at expense of renewables (hydroelectric, wind, solar, geothermal) is not supported by facts. Largest beneficiaries between 1950 and 2006 from federal energy subsidies have been oil and gas receiving more than half of all federal incentives. Primary subsidy for nuclear energy has been R&D. Evaluating the actual electrical energy produced resulting from government subsidy support shows that wind and solar have cost taxpayers 355mils/kWh, coal 1.53 mils/kWh, nuclear 3.8 mils/kWh and hydro at 5.88 mils/kWh. Average cost of U.S. electrical energy in 2006 was 91 mils/kWh so renewables were subsidized at four times the average cost of electricity. Subsidy for Solar Photovoltaic to produce 0.01% of U.S. electricity as of 2006 was $4.43/kWh.

Topics: Governments
Commentary by Dr. Valentin Fuster
2010;():355-359. doi:10.1115/ICONE18-29340.

In Japan, the implementation of the high-level radioactive waste (HLW) disposal is one of urgent issues in the situation that Japan will continue the use of nuclear power. But, the lay people may not have the sufficient amount of information and knowledge about HLW disposal to hold their opinions about this issue. In this research, in order to clarify what opinions they will have with enough information and knowledge, we had the face-to-face dialogues about the HLW disposal with 2 or 3 lay persons. The dialogues were conducted 11 times with different lay persons’ groups. In these dialogues, after the lay participants had a certain amount of knowledge about HLW disposal, they became to talk about their opinions to the HLW disposal program in Japan. These opinions included the doubt against the open solicitation to select the siting area in the HLW disposal program of Japan, the emotion like NIMBY, the indication of lack of public relations about HLW disposal, and so on.

Commentary by Dr. Valentin Fuster
2010;():361-366. doi:10.1115/ICONE18-29416.

This paper proposes and explaines the economic model of MOX fuel used in PWR versus the one used in FBR, considering the fuel flow and the cash flow of fuel cycle. We take the point of view of a firm that is deciding whether to build a new PWR or a FBR, according to its fuel cost and annual cost, ignoring the capital cost and other costs. In order to make the results more practicable, in the future, the economic analysis will provide three sets of costs, which are the best, central and worst cases for the two reactors. The corresponding results, provide analysis of the sensitivity of the costs versus various parameters and other contents.

Commentary by Dr. Valentin Fuster
2010;():367-379. doi:10.1115/ICONE18-29460.

A traditional investment evaluation approach is generally closed with a financial lifecycle performances investigation based on multiple analysis of Discounted Cash Flows (DCF). The international literature is rich of studies about the economics of new Nuclear Power Plants (NPPs), considering the classical accounts related to Construction, Operation & Maintenance, Fuel and Decommissioning. Financial analyses are important but the evaluation of such projects needs a multidimensional approach: besides economics, other technical, social and market factors have to be taken into account too. The Integrated model for the Competitiveness Assessment of SMRs (INCAS), developed by “Politecnico di Milano” cooperating with the IAEA, is designed to analyze the choice of the better NPP size as a multidimensional problem. The INCAS model aims to become the framework for the comparison between “Deliberately Small Medium Reactors (SMRs)” and “Large Reactors (LRs)”, with respect to a specific country situation or “scenario”. In particular the INCAS’s module “External Factors” evaluates the impact of factors not considered in the traditional DCF methods (siting and grid constraints, impact on the national industrial system, etc[[ellipsis]]) but critical for the decision of the plant size to deploy. This paper presents a completely updated version of the “External Factors model” framework under development since 2007. First it presents each factor, providing an adequate background and the quantification procedure, and then each factor is quantified respect to the Italian case. The IRIS reactor has been chosen as SMR representative. Even if the results are related to the Italian situation, they can apply to most of the European countries and the framework of the model can be used for all the countries. The results show that SMRs have better performances than LRs with respect to the external factors, in general and in the Italian scenario in particular.

Commentary by Dr. Valentin Fuster
2010;():381-384. doi:10.1115/ICONE18-29781.

Nuclear power is a safe, clean and economic energy source. The growth of the nuclear power option is impeded in many countries by public concerns over the safety and environmental consequences of producing electricity by means of nuclear reactors. Nuclear power is more compatible with the environment through reduction in emission of green-house gases, fuel diversification, and energy security. Public concern has been expressed in most countries about the construction and operation of nuclear power plants, and this public concern has in many cases led to postponement or failure to start or expand nuclear power programs, and in some cases even caused a retrenchment of existing programs. This paper examines the nature and causes of public concerns about the development nuclear power and the need for public understanding and acceptance of nuclear energy. Some preliminary results on public opinion survey on nuclear energy in Bangladesh are presented in this report. Preliminary survey shows that, Bangladeshi people have a quite satisfactory rate of support to nuclear energy development, which exceeds 60%.

Topics: Nuclear power
Commentary by Dr. Valentin Fuster
2010;():385-390. doi:10.1115/ICONE18-29798.

Being a third generation advanced nuclear plant type, AP1000 has the characteristics of modular construction, “Open-Top” method lifting, etc, which simultaneously present a new challenge for site HSE management. Through studying Health, Safety & Environment (HSE) site management of the first AP1000 nuclear plant (ANP) under construction, this research analyzes the difference of HSE management aspects, such as management commitment and responsibility, HSE awareness and concept, training and education, HSE reward & punishment system and other aspects between China and the west. It also puts forward the specific idea for developing site HSE management system suitable for ANP technology characteristic and China’s actual conditions, which will provide a guarantee for the safe and smooth construction of ANP project.

Commentary by Dr. Valentin Fuster
2010;():391-401. doi:10.1115/ICONE18-29824.

Nuclear Engineering Education has seen a recent surge in activity in the past 10 years in Canada due in part to a Nuclear Renaissance. The Nuclear Industry workforce is also aging significantly and requires a significant turnover of staff due to the expected retirements in the next few years. The end result is that more students need to be prepared for work in all aspects of the Nuclear Industry. The traditional training model used for nuclear engineering education has been an option in an existing undergraduate program such as Chemical Engineering, Engineering Physics, or Mechanical Engineering with advanced training in graduate school. The education model was mostly lecture style with a small number of experimental laboratories due to the small number of research reactors that could be used for experimentation. While the traditional education model has worked well in the past, there are significantly more advanced technologies available today that can be used to enhance learning in the classroom. Most of the advancement in nuclear education learning has been through the use of computers and simulation related tasks. These have included use of industry codes, or simpler tools for analysis of the complex models used in the Nuclear Industry. While effective, these tools address the analytical portion of the program and do not address many of the other skills needed for nuclear engineers. In this work, a set of tools are examined that can be used to augment or replace the traditional lecture method. These tools are Mediasite, Adobe Connect, Elluminate, and Camtasia. All four tools have recording capabilities that allow the students to experience the exchange of information in different ways. The students now have more options in how they obtain and share information. Students can receive information in class, review it later at home or while in transit, or view/participate it live at a remote location. These different options allow for more flexibility in delivery of material. The purpose of this paper is to compare recent experiences with each of these tools in providing Nuclear Engineering Education and to determine the various constraints and impacts on delivery.

Commentary by Dr. Valentin Fuster
2010;():403-408. doi:10.1115/ICONE18-29841.

The results of surveys conducted in some developed countries (the United States, Germany, and France) and in Japan showed that approximately 50% of the respondents considered nuclear power generation to be a cause of global warming. Therefore, it is important to investigate why a wide range of people lack the awareness that nuclear power generation is an effective means of preventing global warming and why approximately 50% of people think that nuclear power generation is a cause of global warming. In this research, it was investigated why people think that nuclear power generation is a cause of global warming. Factor analysis method was applied to data obtained from survey at Kansai area in Japan. Using the survey results, people’s awareness structure was analyzed to determine factors behind people’s perception that nuclear power generation is a cause of global warming and to identify ideas preventing people from recognizing nuclear power generation, which emits no carbon dioxide during power generation, as a means of preventing global warming. As a result, the misunderstanding, the thermal discharge and radioactive material etc. produced from a nuclear power plant promotes global warming, has influenced on this issue. It has become evident that behind such misunderstanding is a negative image of nuclear power. This negative image is a factor to decrease the evaluation that nuclear power is useful for preventing global warming regardless of the presence of the misunderstanding. It is believed that the negative image of nuclear power does not lead to direct association of nuclear power generation and global warming, but by the fear that the accident of the nuclear plant brings the environmental destruction, people evaluate that nuclear power generation is not effective for preventing global warming without grounds. Especially, the tendency is very strong in young people.

Commentary by Dr. Valentin Fuster
2010;():409-414. doi:10.1115/ICONE18-29883.

The high level radioactive waste was decided to the geological disposal in Japan. In addition, operating body and capital management body were decided. However, the place in the repository site has not been decided yet. The selection process in the waste repository is already advanced in Finland and Sweden. In England, some municipalities have expressed interest in HLW. The word used in publicity materials is analyzed, systems in each country are compared, and some good idea for Japanese HLW repository site selection process is studied.

Commentary by Dr. Valentin Fuster
2010;():415-420. doi:10.1115/ICONE18-29983.

The new situations were analyzed for nuclear development and security. Besides nuclear wars and terrorism, reactor runaway and un-ruled radioactive source are the main nuclear accidents in peacetime. With the active development of the nuclear cause in China, the nuclear education system is becoming deficient or defective. In order to ensure the sustainable and efficient accord development of nuclear industry, general nuclear education is necessary for correlative non-nuclear professionals, such as technologies, management, safety, culture and ethic.

Commentary by Dr. Valentin Fuster
2010;():421-424. doi:10.1115/ICONE18-30111.

In order to probe into the usage of the Recommendations of the ICRP, through comparative analysis of low-dose-rate radiation-induced stochastic effects of a nominal risk coefficient, radiation weighting factor, tissue weighting factor as well as the the implementation of changes on the radiological protection system, analysis of the international on Radiological Protection fundamental recommendations of the Committee on the latest changes in radiological protection and development, and that these changes can not affect the existing radiation protection of China’s basic policy and standards.

Commentary by Dr. Valentin Fuster
2010;():425-430. doi:10.1115/ICONE18-30215.

RELAP/SCDAPSIM, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). RELSIM, an advanced interactive simulator Graphical User Interface, is a commercially available package being developed by Risk Management Associates (RMA). The combined package, RELSIM-RELAP/SCDAPSIM, is being used for training of university students and other novice reactor systems analysts to help them understand how complex thermal hydraulic and/or reactor systems perform under realistic and postulated conditions. RELAP/SCDAPSIM uses internationally developed and validated system thermal hydraulic, fuel behavior, and severe accident models in combination with a flexible building block approach to model thermal hydraulic and reactor systems. RELSIM also uses a building block approach and user defined graphics screens in combination with the ability to interactively control the RELAP/SCDAPSIM simulation. As a result, students and analysts can use the package to describe both simple and complex thermal hydraulic systems ranging from simple pipes and university-scale experimental facilities up through current and conceptual reactor systems. This paper gives a brief description of the RELSIM-RELAP/SCDAPSIM package and then provides a discussion of how the package is being used for student and analyst training, The discussion includes the development of training tutorials and videos for novice users, the development of representative sample problems and graphics displays, and the type of training provided to support the users. A case study is presented in the paper outlining the sample problems and displays for use in novice RELAP5 and RELAP/SCDAPSIM user training classes.

Commentary by Dr. Valentin Fuster

Student Paper Competition

2010;():431-445. doi:10.1115/ICONE18-29007.

For nuclear power plant, probabilistic safety assessment (PSA) is an effective tool for risk evaluating, risk recognizing and risk managing. This paper initiate a study focused on shutdown safety for NPP as well as the related PSA approaches. Especially, a PSA preliminary analysis and estimation on shutdown operation to the 1000 MWe NPP are performed. The first part of this thesis presents the methods used in the classification of plant operation state (POS), the identification of initiating event and the quantification of initiating event frequency. The second part centers on configuration of shutdown model used in PSA. Based on the design and operation features during shutdown conditions of NPPs, event trees for the shutdown model are constructed and Human Reliability Analysis (HRA) is performed. The last part is the quantification of the integrated shutdown model by means of Risk Spectrum software. System analysis is based on FuQing nuclear power plant PSA report. Component failure data and initiating event frequencies are largely based on EPS900 or NUREG./CR-6144. Overall point estimation of core damage frequency (CDF) during shutdown is assessed; the most important risk contributors of plant operational states, the dominant initiating events and their contribution to CDF are described respectively.

Topics: Safety
Commentary by Dr. Valentin Fuster
2010;():447-456. doi:10.1115/ICONE18-29099.

This work gives the theoretical description with a practical measurement of neutron and gamma spatial distributions in a radial experimental channel (RC) at VR1 research reactor. For providing the measurements an “off-line” and “on-line” methods were used. The “on-line” measurement of neutron spatial distribution was performed with a 3 He gas filled detector. From the “off-line” measurements Neutron Activation method (for neutron detection) and Thermoluminescent method (for gamma radiation detection) were chosen. Results obtained from experiments were compared with the results from MCNP code.

Commentary by Dr. Valentin Fuster
2010;():457-462. doi:10.1115/ICONE18-29173.

Adequate design engineering and maintenance of circuits with fast neutron reactors cooled with lead and lead-bismuth coolants require considering the peculiarities of hydrodynamics of these coolant flows. It is traditionally reputed that the hydrodynamic characteristics of heavy liquid-metal melts are analogous to the characteristics of water and primary sodium, which is practically valid for the conditions of part of the equipment and channels of a reactor circuit. The main peculiarities of heavy liquid-metal coolants compared to water and primary sodium, which affect the flow characteristics, are: - unwettability of channels with oxide protective coatings of reactor circuits by lead and lead-bismuth eutectic melts; - high boiling temperature exceeding the fusion temperature of steel; - high density exceeding by an order the densities of water and natrium; - low solubility of impurities in lead and lead-bismuth eutectic melts; - higher surface tension coefficient. The design value of saturated vapors of lead and its alloys at the temperatures 400–550 °C is 10−18 –10−10 at (1 at = 0.1 MPa), which is essentially less than the values of natrium and water. Processes of traditional cavitation in the flow of heavy liquid-metal coolants cannot occur because of their specific character. The main circulation pumps are a basic element of reactor circuits. In fact, the flow sections of these pumps and those of other vane-type pumps operating in lead and its alloys cannot be calculated by traditional methods as far as cavitation characteristics are concerned; adequate calculation formulas are not available now. In a channel with walls unwettable by a flow of heavy liquid metal, this flow contacts with walls by means of the boundary layer having specific properties (surface energy, etc.) analogous to those of free surfaces of melts contacting with gas. Internal pressure in the flow forces liquid metal against walls, thus the liquid metal speed in the region of their contact is zero. As the pressure in the flow decreases due to growth of speed or other effects, the outer layer of the liquid metal flow can move away from the wall; in this case water appears on its surface. To study cavitation processes in a heavy liquid-metal coolant flow, the authors have carried out the following experiments: - determining the conditions of disconnection of liquid lead and lead-bismuth eutectic column; - determining the cavitation characteristics of the centrifugal pump pumping lead at the temperature 500 °C; - comparative investigation of the characteristics of Venturi nozzle in water and liquid metal. The experimental study of the characteristics of disconnection of heavy liquid-metal coolant column has shown that disconnection occurs at the boundary of liquid and cold metals; the reason of disconnection is leakage of gas from melt volume and, perhaps, from the near-wall region; disconnection occurs at negative voltages in the cross section of the column. The experimental study of the cavitation characteristics of the centrifugal pump at the temperature of pumped lead 500 °C and the circumferential speed of about 15 m/s has show that failure (cease) of pumping takes place at the pressure at the impeller inlet of about 19.6–24.5 kPa. Continuous operation of pump in the regime of pumping failure does not lead to destruction of the flow part surfaces of the pump. The character of the process corresponds to the so-called gas cavitation and is completely inconsistent with traditional cavitation. The experimental comparative study of the hydrodynamic characteristics of the same Venturi nozzle for water current at the temperature T = 20 °C and lead-bismuth eutectic at T = 350 °C without gas supply and with gas supply at the speeds 10–20 m/s has shown the following. The hydraulic resistance of the eutectic nozzle is more than an order higher than the analogous value for water under the same test conditions. This is, probably, due to flow disconnection and jet contraction in the narrow part of the nozzle with formation of water on its surface and backflows in the nozzle diffuser. Supply of relatively small amounts of gas into the narrow part slightly varies the characteristics of the processes. The consideration of specific character of heavy coolant flow hydrodynamics is required for adequate design engineering and maintenance of some elements of reactor circuit.

Commentary by Dr. Valentin Fuster
2010;():463-469. doi:10.1115/ICONE18-29175.

The increasing interest to reactors of heavy nuclear fission by fast neutrons initiates the research substantiating the operational integrity of mechanisms with contact surfaces in a medium of high-temperature heavy liquid-metal coolants.

Commentary by Dr. Valentin Fuster
2010;():471-479. doi:10.1115/ICONE18-29198.

Understanding the freezing behavior of molten metal in flow channels is of importance for severe accident analysis of liquid metal reactors. In order to simulate its fundamental behavior, a 3D fluid dynamics code was developed using Finite Volume Particle (FVP) method, which is one of the moving particle methods. This method, which is fully Lagrangian particle method, assumes that each moving particle occupies certain volume. The governing equations that determine the phase change process are solved by discretizing its gradient and Laplacian terms with the moving particles. The motions of each particle and heat transfer between particles are calculated through interaction with its neighboring particles. A series of experiments for fundamental freezing behavior of molten metal during penetration on to a metal structure was also performed to provide data for the validation of the developed code. The comparison between simulation and experimental results indicates that the present 3D code using the FVP method can successfully reproduce the observed freezing process such as molten metal temperature profile, frozen molten metal shape and its penetration length on the metal structure.

Commentary by Dr. Valentin Fuster
2010;():481-485. doi:10.1115/ICONE18-29253.

In this study, the multiplication factor and neutron spectrum behaviors were investigated against the moderator-to-fuel ratio, the fuel loading height, and the detector location in HTR-10. The MCNP5 computer code (version 1.51) was employed to perform all the simulation computations. The results revealed that the multiplication factor varies significantly depending on the moderator-to-fuel ratio and the fuel loading height due to the competition among the neutron moderation and absorption abilities of the moderator as well as the neutron production ability of the fuel. Due to its inherent stability, HTR-10 is deliberately designed such that the multiplication factor decreases and the neutron spectrum softens as the moderator-to-fuel ratio increases. The average neutron energy level in the HTR-10 fuel balls is approximately 200 keV and ranges from smallest to largest at the middle, bottom, and top of the reactor core, respectively.

Topics: Fuels , Design
Commentary by Dr. Valentin Fuster
2010;():487-495. doi:10.1115/ICONE18-29283.

Different people have different opinion on nuclear power use. As for HLW disposal, more complex issues such as uncertainty due to very long time-scale required for disposal would add to the difficulty of conducting smooth communication and building stable consensus between the scientific experts and the general public. We carried out Q&A dialogue experiments between an expert and public participants who were indifferent to HLW disposal. The dialogue process was divided into two periods. In the learning period, participants learned fundamental knowledge about HLW disposal. In the discussion period, participants discussed about one specific topic, such as “Resource and Energy”, “Geological Disposal”, or Safety Violation. These dialogue experiments can help the experts to communicate and conduct comprehensive activities with the public which would help implement the HLW disposal from the public point of view.

Commentary by Dr. Valentin Fuster
2010;():497-505. doi:10.1115/ICONE18-29297.

A multi-channel model thermal-hydraulic analysis code in real-time for plate type fuel reactor is developed in this paper. In this code, every fuel assembly in reactor is divided into a subchannel. A series of reasonable mathematical and physical model are set up based on the structure and operational characteristics of plate type fuel core. As for the choice of flow friction and heat transfer models, all possible flow regimes which include the laminar flow, transient flow and turbulent flow, and heat transfer regimes which include single liquid phase heat transfer, sub-cooled boiling, saturation boiling, film boiling and single vapor phase heat transfer, are considered. The correlations and constitutive equations used in the code are fit for the rectangular channel. Look-up table method is used to calculate the properties of water and steam. The code has been loaded on the real-time simulation supporting system SimExec. The reactivity insertion accident and loss of flow accident, which has been defined in the IAEA 10MW MTR benchmark program, were calculated by the code in this paper for validation. Furthermore, the steady state of CARR (China Advanced Research Reactor) is analyzed by this code. The detailed flow distribution in each fuel assembly is obtained. The temperature of coolant, quality, void fraction, DNBR in each subchannel is calculated. The results show that the recently developed code can be used for real time thermal hydraulic analysis of plate type fuel reactor.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2010;():507-516. doi:10.1115/ICONE18-29321.

Probabilistic Safety Assessment (PSA) has been an important tool to assist Nuclear Power Plants (NPPs) management over the years. Through PSA, the weak links of the whole system can be identified by component importance measures. The importance measures can be classified according to risk significance and safety significance. They signify the role that each component plays in either causing or contributing to the occurrence of an undesired event. A wide range of importance measures have been developed over the years and most of them are geared towards coherent systems. Importance analysis of non-coherent system is rather limited. In this paper, a set of component importance measures for non-coherent systems is analyzed and investigated. The comparison and the selection of the most informative and appropriate measures for guiding the maintenance of a system are presented. A steam generator level control system of NPP is used to demonstrate the application of the results.

Topics: Maintenance
Commentary by Dr. Valentin Fuster
2010;():517-525. doi:10.1115/ICONE18-29344.

High temperature gas-cooled reactor with direct helium turbine cycle is based on the closed Brayton cycle. Its outstanding feature is the high efficiency of power generation. Pervious researches showed that recuperator was the key component to promote the cycle’s efficiency. And pressure drops in components were unavoidable in actual projects and had significant influence on cycle efficiency. A dimensionless model was proposed to analyze cycle’s features of HTGR coupled with gas turbine. The parameters’ effect on cycle’s efficiency was analyzed, with full consideration of the frictional and local pressure drops respectively. Under the restriction of materials and state-of-art of technologies, it showed that the cycle’s efficiency depended on compression ratio, recuperator’s effectiveness and pressure drops of components. However the pressure drop ratios of different components were inherently connected due to the closed cycle. Furthermore pressure drops inside the recuperator were also the function of effectiveness of the heat transfer based on the Reynolds analogy. Therefore cycle’s efficiency just depended on recuperator’s effectiveness with fixed compression ratio. So there existed optimal recuperator’s effectiveness and maximum cycle’s efficiency, which varied with the pressure ratio and other parameters as temperature ratio. The calculation also indicated that the pressure drop in pipes was close to that in heat exchangers. That was, the local pressure drop and frictional pressure drop should be considered respectively, and the local pressure drop made quite large reduction of cycle’s efficiency. The result also showed that local pressure drop had great influence on parameters such as optimal compression ratio and recuperator’s effectiveness.

Commentary by Dr. Valentin Fuster
2010;():527-532. doi:10.1115/ICONE18-29388.

The LMFBRs (L iquid M etal Cooled F ast B reading R eactors) is a key technology for the future electricity demand. Therefore, the establishment of safety of LMFBRs is deeply desired. To evaluate the safety of LMFBR steam generators, empirical studies have been performed for the ruptures of heat transfer tube caused by the overheat due to sodium-water reaction in case of a practical scale conditions. But few systematical experiments have been performed for the clarification of phenomena from the viewpoint of thermal hydraulics and physical chemistry. The absence of such studies is derived from the fact that sodium is chemically active and is not feasible for visualized experiments. To evaluate the safety against the secondary failure of heat transfer tubes by an analytical code, which has been tried in JAEA, it is required to understand the heat transfer phenomena around the tube. In this study, we investigated experimentally the thermal hydraulics behavior around a single heated rod with sodium-water reaction as an essential study for the clarification of raptures phenomena of heat transfer tube. The experimental apparatus was consisted of a sodium pool tank, an electrically heated test rod, a gas jet nozzle of argon-water mixture. We set horizontally in the sodium pool the test rod that is heated with the constant heat flux, and to which a gas jet of argon-water mixture was supplied for the sodium-water reaction from its lower side. The temperature of sodium pool was kept to be 420K so that a product of sodium-water reaction, NaOH (melting point is 591K), may exist as a solid phase. Gas jet velocity was set to be 17.3m/s, and the amount of water vapor in the gas mixture was 3% in mass. Just after the introduction of gas mixture, the temperature of sodium pool increased by the heat of chemical reaction. At the same time, the heat transfer to the rod surface decreased rapidly. By the observation after the experiment, it was confirmed that coarse reaction products deposited thickly at the upper side of the rod and finely granular products adhered to the lower ones. Thus, at low-sodium temperature conditions, the products of sodium-water reaction on the rod surface cause the decrease of heat transfer rate between the rod and sodium pool, which depends on the local distribution of deposits. The present authors therefore obtained experimentally the phenomena important for the development of an analytical code of JAEA.

Commentary by Dr. Valentin Fuster
2010;():533-536. doi:10.1115/ICONE18-29408.

This paper reviews some moderators for cold neutron source. As a good cold moderator, solid methane was studied and evaluated using the new synthetic frequency spectrum theory (SFS). Due to its high proton density and easy handling, mesitylene (not included in ENDF/B) has also been considered as a very good moderator for cold neutron source. Evaluation of this material in different crystalline phases was done by using a preliminary frequency spectra built combining experimental and synthetic contributions. As a result, we have generated ACE format scattering data files by the NJOY code, with validations of comparing total cross sections from experiments.

Commentary by Dr. Valentin Fuster
2010;():537-542. doi:10.1115/ICONE18-29421.

There are several methods for heat transfer enhancement. For example, there are attaching various fins on the heat transfer surface, processing the surface roughly, inserting twisted tape, and so on. These methods increase heat transfer coefficient or area by manufacturing of the heat transfer surface. However, it has to take into consideration the deterioration of the structure strength by attaching the fins on the tube surface with the design of the heat exchanger. The objective of this study is to clarify characteristics of heat transfer and pressure drop in the channel inserted metallic wire with high porosity. A heat transfer experiment has been performed using a horizontal circular tube to obtain the heat transfer characteristics in the channel inserted copper wire. This paper describes the heat transfer and flow characteristics of a heat exchanger tube filled with a high porous material. Fine copper wire (diameter: 0.5 mm) was inserted in a circular tube dominated by thermal conduction and forced convection. Working fluid was air. Hydraulic equivalent diameter was cited as the characteristic length in Nusselt number and Reynolds number. From the results obtained in this experiment, it was found that an amount of heat transfer in the tube with the copper wire was larger than that without one. An effectiveness of heat transfer enhancement increased with the temperature of the heated wall. The amount of heat transfer in the circular tube inserted copper wire, which has 0.993–0.998 of porosity, increased about 15% comparing with the tube having a smooth wall surface under the condition of the constant heat flux and lower than 170°C of the wall temperature.

Commentary by Dr. Valentin Fuster
2010;():543-549. doi:10.1115/ICONE18-29437.

This paper describes the development of a compact engineering simulator of Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM) by embedding THERMIX code into the vPower simulation environment. The compact engineering simulator consists of modules for two reactors, two steam generators and entire secondary loop system for power generation with a water-steam Rankin cycle. Two THERMIX modules are employed to simulate the two primary loops corresponding to the two reactors in the HTR-PM respectively. Then, the vPower synchronizes the two THERMIX modules in executing simulation to such two reactor module-structures of HTR-PM. The simulation modules for secondary loop and human machine interface are mainly developed with intrinsic models of vPower simulation platform. Current simulation results are in good agreement with the design values and safety analysis results of HTR-PM. The compact HTR-PM engineering simulator will be improved and validated for safe analysis, procedure development and design change verification.

Topics: Simulation
Commentary by Dr. Valentin Fuster
2010;():551-555. doi:10.1115/ICONE18-29447.

An exponential function expansion nodal diffusion method is proposed to take care of diffusion calculation in unstructured geometry. Transverse integral technique is widely used in nodal method in regular geometry, such as rectangular and hexagonal, while improper in arbitrary triangular geometry because of the mathematical singularity. In this paper, nodal response matrix is derived by expanding detailed nodal flux distribution into a sum of exponential functions, and nodal balance equation can be obtained by strict integral in the polygonal node. Numerical results illustrate that the exponential function expansion nodal method in rectangular and triangular block can solve neutron diffusion equation in regular and irregular geometry.

Commentary by Dr. Valentin Fuster
2010;():557-564. doi:10.1115/ICONE18-29463.

This paper aims to study the pressure distribution and flow patterns in the top fuel region of the AP1000™ reactor using CFD. This study is being performed as part of a CFD evaluation of the flow in the top fuel and upper plenum regions of a PWR reactor vessel. The flow patterns, including cross flows in the top fuel region, are inter-related with the flow distribution and pressure forces in the reactor vessel upper plenum region. Before detailed computations of the flow in the whole top fuel and upper plenum region are performed, conducting local computations for segments of the domain can provide information about physical aspects of the flow as well as mesh sensitivities. The domain of interest in this paper is the top fuel region including the upper part of the fuel assembly (top grid, fuel rods, top nozzle), upper core plate, and core component hold-down device. The commercial CFD computer code STAR-CCM+ is employed to generate the computational mesh, to solve the Reynolds-averaged Navier-Stokes equations for incompressible flow with a Realizable k-epsilon turbulence model, and to post-process the results. The complicated geometry of the top fuel region needs to be simplified so that the mesh size for the CFD model of the whole upper plenum and top fuel region does not exceed current software and hardware capabilities. In this study, several different trimmed meshes have been generated to study the effects of the geometries of the hold-down device and the lateral flows. Mesh sensitivity studies have been conducted for each individual part, i.e., the top grid, top nozzle, upper core plate, and hold-down device, in order to determine the proper geometrical simplifications. Pressure drop measurement data are compared with the predicted CFD results and act as a guideline for the mesh selection. These studies support the applicability of the geometrically simplified models and chosen mesh size for the CFD model of the full upper plenum and top fuel regions.

Commentary by Dr. Valentin Fuster
2010;():565-571. doi:10.1115/ICONE18-29517.

Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase simulation method such as interface tracking method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as rod bundles. However, since two-fluid model include a lot of constitutive equations, applicability of these constitutive equations must be verified by use of experimental results, and the two-fluid model has problems the result of analyses depends on accuracy of constitutive equations. To solve these problems, an advanced two-fluid model has been developed in Japan Atomic Energy Agency. In the model, an interface tracking method is combined with the two-fluid model to predict large interface structure behavior accurately. Liquid clusters and bubbles larger than a computational cell are calculated using the interface tracking method, and those smaller than a cell are simulated by the two-fluid model. Constitutive equations to evaluate the effect of small bubbles or droplets on two-phase flow required in the advanced two-fluid model as same as a conventional two-fluid model. However, dependency of small bubbles and droplets on two-phase flow characteristic is relatively small, and the experimental results to verify the equations are not required much. The turbulent dispersion force term is one of the most important constitutive equations for the advanced two-fluid model. The turbulent dispersion force term has been modeled by many researchers for the conventional two-fluid model. However, the existing models include effects of large bubbles and deformation of bubbles implicitly, these models are not applicable to the advanced two-fluid model. In this study, we develop the new model for turbulent dispersion force term. In this model, effect of large bubbles and deformation of bubbles are neglected. The liquid phase turbulent kinetic energy and bubble-induced turbulent kinetic energy are considered to evaluate driving force in the turbulent diffusion of small bubbles. The bubble-induced turbulent kinetic energy is given by the function of bubble diameter and local relative velocity, and the liquid phase turbulent kinetic energy is similar to the single phase flow case. Furthermore, we considered energy transfer from the bubble-induced kinetic energy to the liquid phase turbulent kinetic energy. To verify the developed model, the advanced two-fluid model and the model for turbulent dispersion term were incorporated to the 3-dimensional two-fluid model code ACE-3D, and comparisons between the results of analyses and air-water two-phase flow experiments in 200 mm diameter vertical pipe were performed.

Commentary by Dr. Valentin Fuster
2010;():573-579. doi:10.1115/ICONE18-29545.

This paper presents experimental data of cavitation experiments to determine the characteristics of cavitation and erosion in the flowing liquid sodium at 200–400°C. The test section is a venturi made from SUS316 since this material is used as the cladding material for SFR. The ID and OD of the venturi test section are 6.5 mm and 21.4 mm, respectively. The data show that the onset cavitation conditions (onset velocities) in liquid sodium are influenced by the change of the stagnant pressure at the expansion tank. Cavitation noise signals at developed cavitation conditions are not affected by the change of stagnant pressure and might be caused by choking of the sodium flow that restricts the formation of cavitation bubbles. For all the different stagnant pressures and temperatures, the onset cavitation coefficient K is around unity. Meanwhile, erosion experiment in the flowing liquid sodium for 600 hours at 200°C and cavitation coefficient value K of 0.59–0.51 (developed cavitation condition) reveals that cavity bubbles produce some micro pits at the outlet of the venturi test section. These pits indicate that erosion is occurred on the surface of the tested material and could be a major problem for the development of SFR if cavitation occurs inside the critical/important parts.

Commentary by Dr. Valentin Fuster
2010;():581-586. doi:10.1115/ICONE18-29546.

On the premise that ensuring the reliability of structural strength, and in order to get the optimal structural parameters and the number of reinforcement rib of the ITER feeder S-bend box (SBB), so that its stress distribution is more uniform and reasonable under the critical pressure load, and the cost of materials is relatively smaller, this article, through theoretical calculations and inferences on the basis of the parametric modeling and static analysis and checking of SBB, setting up different reinforcement rib numbers, optimizes through the ANSYS optimization design module, with the maximum stress as the objection function, with and reasonable quality and displacement as state variables, with the size of wall thickness and the space of reinforcement rib as design variables. And it regulates true stress at stress concentration through submodeling of ANSYS and local solid modeling. The results of optimization analysis show that: SBB reached the optimal solution when the reinforcement rib number N of SBB takes 3, the maximum equivalent stress is 117 Mpa and the weight is 6417Kg. Finally, SBB structural parameters, which obtained through optimization design, are rounded according to GB. These meet the design requirements, correspond to the practical applications and provide technical parameters and basis for the future development of SBB.

Commentary by Dr. Valentin Fuster
2010;():587-598. doi:10.1115/ICONE18-29550.

Two-phase flow instability and dynamics of a parallel multichannel system has been theoretically studied under periodic excitation induced by rolling motion in the present research. Based on the homogeneous flow model considering the rolling motion, the parallel multichannel model and system control equations are established by using the control volume integrating method. Gear method is used to solve the system control equations. The influences of the inlet, upward sections, and heating power on the flow instability under rolling motion have been analyzed. The marginal stability boundary (MSB) under rolling motion condition is obtained. The unstable regions occur in both low and high equilibrium quality and inlet subcooling regions. The multiplied period phenomenon occurs in the high equilibrium quality region and the chaos phenomenon appears on the right of MSB. The concept of stability space is presented.

Commentary by Dr. Valentin Fuster
2010;():599-608. doi:10.1115/ICONE18-29568.

Supersonic steam injector is a passive jet pump which operates without rotating power source or machinery and it has high heat-transfer performance due to the direct contact condensation between supersonic steam flow and subcooled water jet at the mixing nozzle. Since the supersonic steam injector has a quite simple and compact structure, it has been considered to apply to the safety system for the Next-generation nuclear power plant. There are various researches about the formulation and modeling of operating, flow structure and heat transfer characteristics of both vapor and liquid flow. However, there are few models which are capable of evaluating heat and momentum exchange at the boundary layer between supersonic steam flow and water jet. Since heat and momentum exchange is considered to have a major impact to operating characteristics of a supersonic steam injector, it is necessary to formulate the model which simulates such complex phenomena at the boundary layer with high accuracy. The objective of the present study is to investigate the relation between the thermal characteristics and interfacial behavior between the flows to develop a model which is able to assess the heat and momentum transfer characteristic and the flow structure of the supersonic steam injector in detail. In the present study, a visible test section of water jet-centered supersonic steam injector was adopted to conduct visualization of the water jet with high speed video camera. In addition, special measurement instrumentations of temperature and pressure were applied to obtain radial distribution of temperature and pressure in the mixing nozzle of the injector. There were large velocity and temperature gaps between the water jet and the supersonic steam flow which indicated the existence of large momentum and heat exchange at the boundary layer of the flows. It was clarified that there was pressure gradient which was considered to stabilize the water jet in the mixing nozzle from calculation of radial distribution of total pressure gradient. From the visualization measurement, it was also clarified the existence of a complex wavy behavior on the surface of the water jet. The wave velocity was estimated by the image processing technique and the cross-correlation method. It was found that there was a relation between the wave velocity and heat transfer characteristics in the supersonic steam injector. It is suggested that the enthalpy ratio between the liquid subcool enthalpy and steam condensation enthalpy as well as the Jacob number between both flows could be an indication factor for the effect of the wavy behavior on the condensation.

Commentary by Dr. Valentin Fuster
2010;():609-614. doi:10.1115/ICONE18-29572.

The resistance characteristics of air-water flow upward through packed bed have been studied experimentally. Experiments were conducted in transparent tube with 50mm inner diameter filled with glass spheres of which the diameters are 2, 5 and 8mm, respectively. Experimental results show that the pressure drop increases with the increasing of gas-liquid mass flow rate, and has a certain relationship with the flow pattern. The different particle diameter and porosity have great influence on pressure drop under the same flow condition. The applicability of several representative correlations for calculating the two-phase pressure drop and two new correlations were evaluated against 234 group experimental data. All compared correlations can be grouped into two, namely: (a) the separated flow model based on Lockhart & Matinelli method (b) the homogeneous model based on gas and liquid Reynolds number. The results show that: (1) the best agreement between measured and calculated values is obtained with the correlations based on separate flow model, but the predictive ability is reduced with the increasing of particle diameter. (2) the existing homogeneous model shows the considerable discrepancies with experiment values; however the modified homogeneous correlation by considering particle-to-column diameter ratio and porosity gives good agreement with experimental data.

Topics: Pressure drop , Water
Commentary by Dr. Valentin Fuster
2010;():615-621. doi:10.1115/ICONE18-29585.

During the severe accident in a nuclear power plant, large amounts of fission products release with accident progression, which includes in-vessel release and ex-vessel release. Mitigation of release of fission products is the need of alleviating radiological consequence in severe accident. Mitigation countermeasures to in-vessel release of fission products are studied, including feed-bleed in primary loop, feed-bleed in secondary loop and cooling of ex-vessel. Representative high pressure melt accident of station blackout is chosen, and different entry condition of countermeasures is assumed. The results show that: (1) Feed-bleed in primary loop is an effective countermeasure to mitigate in-vessel release of fission products. With early time to implement the countermeasure, in-vessel release fraction of fission products is low. (2) Feed-bleed in secondary loop is also an effective countermeasure to mitigate in-vessel release of fission products. Low in-vessel release fraction of fission products is produced with early time of countermeasure implemented. (3) Cooling of ex-vessel is not an effective countermeasure to control in-vessel release of fission products, the in-vessel release fraction in this case is almost equal to base case that uses none countermeasure.

Topics: Accidents , Vessels
Commentary by Dr. Valentin Fuster
2010;():623-627. doi:10.1115/ICONE18-29590.

High-level liquid waste (HLLW) generated from reprocessing process contains actinides, lanthanides, fission products (FP) and a significant amount of nitrate ion. The partitioning and transmutation concept has been introduced for reducing the long-term hazards of HLLW. Several chemical separation processes mainly based on solvent extraction methods have been proposed to treat HLLW. However, solids consisting mainly Mo and Zr are known to form in HLLW during its long-term storage, Solid formations influence the composition of HLLW and the downstream solvent extraction process. To understand the precipitation behavior and stability of HLLW during its long-term storage, simulated HLLW (prepared as raffinate solution from LWR spent fuel reprocessing, 1AW solution) was prepared. Preliminary studies on solid formation behaviors with regard to the precipitation formation during refluxing and aging (representing a long-term storage) were carried out. Precipitation kinetics of major FPs such as Zr, Mo, Ru, rare earth elements, and etc. have been studied; The effect of phosphate ion concentration and temperature on solids formation were also experimentally examined. The formation conditions and the mechanism of solids were discussed.

Topics: Solids , Storage
Commentary by Dr. Valentin Fuster
2010;():629-634. doi:10.1115/ICONE18-29612.

Collapse of a water jet flowing out from a nozzle to the atmosphere was examined. The diameters of nozzles used in the experiments were 3, 6 and 8 mm. The flow state of the water jet was recorded with a high speed video camera. The collapse length was derived from recorded images. When the flow velocity was quite low, the surface of the water jet was smooth and small perturbations appeared at the lower position of the water jet. As the flow velocity was increased, the position where the small perturbations appeared came close to the nozzle outlet. The perturbations grew as these went downstream and lumps of water were formed at the lower position. When the flow velocity was further increased, successive waves came around on the surface of the water jet. The collapse of the water jet occurred in such a state that the lump of water was torn off from the jet. When the water jet velocity was high, the jet turned into a dispersed flow and the collapse occurred. The agreement of the measured results and the predicted results was poor. It was considered that the instability of the surface of the water jet seemed important for the jet collapse in the present experimental range. The Kelvin-Helmholtz instability wave length was compared with the measured wave length on the water jet. When the wave length reached the Kelvin-Helmholtz instability wave length, the jet collapse occurred except the case that the transition to the dispersed flow caused the jet collapse.

Commentary by Dr. Valentin Fuster
2010;():635-644. doi:10.1115/ICONE18-29618.

For the safety design of the Fast Breeder Reactor (FBR), the Post Accident Heat Removal (PAHR) is required when a hypothetical Core Disruptive Accident (CDA) occurs. In the PAHR, it is strongly required that the molten core material can be cooled down and solidified by the sodium coolant in the reactor vessel. There is high possibility for molten material to be ejected as a liquid jet into sodium coolant in the reactor vessel. In order to estimate whether the molten material jet is completely solidified by sodium coolant or not, it is necessary to understand the interaction between molten core material and coolant such as jet breakup and fragmentation behavior in coolant. The jet breakup behavior is the phenomenon that the front of molten material breaks up in coolant. To clarify the mechanism of jet breakup and fragmentation during the CDA for the FBR, it is necessary to understand the correlation between jet breakup lengths and size distribution of fragments when molten material jet interacting with coolant. The objective of the present study is to clarify the dominant factor of the jet breakup length and the size distribution of fragments experimentally. Molten jet of U-alloy 138 is injected into water as simulated core material and coolant by free-fall. The density ratio of core material and coolant is almost same as that of the real FBR system. The jet breakup behavior as interaction of molten material with coolant is observed with high speed video camera. Front velocity of the molten material jet is estimated by using the image processing technique. It suddenly decreases when the jet fall into the coolant. The jet breakup length estimated from observed images is compared with the breakup theories to understand the effect of experimental parameters for the jet breakup length. The solidified fragments are gathered and classified in size, and the mass in each size is measured. Median diameter is obtained from the mass distribution of the fragments. In comparison with interfacial instabilities, the median diameter of fragments shows the independent of relative velocity. The jet breakup lengths and median diameters compared with existing theories is discussed.

Commentary by Dr. Valentin Fuster
2010;():645-649. doi:10.1115/ICONE18-29631.

Bubble carry-under into the water pool was examined. In experiments, a water jet from a nozzle of 5 mm in diameter plunged into the water pool. The distance between the nozzle outlet and the pool surface was 246 mm. Flow behavior in the water pool and also the state of the water jet surface were recorded with a high speed video camera. Following conclusions were obtained. When the flow rate of the water jet was small, the water jet disintegrated into small drops on the way from the nozzle outlet to the pool surface. The wave appearing position moved downward as the flow rate was increased. When the wave length reached the Kelvin-Helmholtz critical wave length, the water jet disintegrated into drops. When flow rate of the water jet was increased, the surface of the water jet became smooth and no perturbation was observed. The carry-under was not observed in this situation. When the flow rate of the water jet was further increased, large waves came to appear on the water jet surface. The wave appearing position moved upward as the flow rate was increased. Even if the wave length on the water jet reached the Kelvin-Helmholtz critical wave length, the water jet did not disintegrate into drops and the water jet plunges into the pool with large waves on the water jet. The penetration depth in this case was deep and the volume of the bubble carry-under was large compared with the case that the water jet disintegrated into drops.

Topics: Flow (Dynamics)
Commentary by Dr. Valentin Fuster
2010;():651-656. doi:10.1115/ICONE18-29633.

This study examines a corrosion control technique for corrosion-resistant materials or of stainless steel in piping for nuclear reactors. This employs an effect of Radiation Induced Surface Activation (RISA). The experimental results revealed: (1) The mechanism behind the corrosion control proposed by the previous report was confirmed to be appropriate. This via tests that measured the amount of dissolved oxygen and iron ions, in the solution. (2) The corrosion control technique was confirmed to be useful for stainless steel with any kind of metal oxide film coating on the surface. (3) It was also shown to be useful even in actual seawater, due to biological effects, which is a far more severe environment for corrosion control than simple salt water. The corrosion control technique for corrosion-resistant material using RISA in seawater has therefore been shown to offer a significant potential for practical applications in naval architecture and marine structures.

Commentary by Dr. Valentin Fuster
2010;():657-664. doi:10.1115/ICONE18-29636.

In relation to the development of the interfacial area transport equation, a precise database of the axial development of void fraction profile, interfacial area concentration and Sauter mean bubble diameter in an adiabatic nitrogen-water bubbly flow in a 5 mm-diameter mini pipe was constructed for normal and microgravity conditions using stereo image-processing. The flow measurements were performed at four axial locations (axial distance from the inlet normalized by the pipe diameter, z/D = 5.5, 34, 72 and 110) and with various flows: superficial gas velocity of 0.00434–0.0420 m/s, and superficial liquid velocity of 0.239–0.949 m/s. The effect of gravity on radial distribution of bubbles and the axial development of two-phase flow parameters is discussed in detail based on the obtained database and visual observation.

Commentary by Dr. Valentin Fuster
2010;():665-673. doi:10.1115/ICONE18-29708.

The freezing and penetration of molten core fuel and structural materials penetrating into flow channels are important thermal-hydraulics phenomena to safety assessment of postulated core disruptive accidents in liquid metal reactors. The main objective of this study is to investigate fundamental characteristics of freezing and penetration behavior involved in melt and solid mixture flowing on-to structure material. In our study, solid copper particles mixed with molten wood’s metal (melting point 78.8°C) was used as a simulant melt, while stainless steel and brass were used as freezing structures. A series of fundamental experiments was performed to study the effects of solid particles on the freezing and penetration behavior under the various thermal conditions of molten metal and varying solid particle volume fraction and structure metal. The melt flow and distribution were observed using a digital video camera. The melt penetration length on the structure and proportion of adhered frozen metal on to structure surfaces were measured in the present series of experiments. The results indicate that penetration length becomes shorter with increasing solid particles volume fraction in melt. The present results will be utilized to build a relevant database for verification of fast reactor safety analysis codes.

Commentary by Dr. Valentin Fuster
2010;():675-680. doi:10.1115/ICONE18-29724.

All of the RAFM steels only safely used under 550°C, that is not enough for the next reactor. An new RAFM steel was melted by non-vacuum induction melting (VIM) and electro-slag remelting (ESR), followed by hot-forging and rolling into rods and plates. In this paper, we investigated the effect of thermal ageing treatment on tensile properties of the rods and plates. The microstructure was studied by OM (optics micrograph) and scanning electron microscopy (SEM). The results showed that by using the same heat treatment process, the tensile strength of the samples was 680MPa, the total elongation was 31%, which were better than the CLAM steel whose tensile strength and total elongation were 668MPa and 25% respectively. The difference between the transverse and the longitudinal properties was reduced markedly. So the ESR played an important part in improving the mechanical properties.

Topics: Steel , Melting
Commentary by Dr. Valentin Fuster
2010;():681-691. doi:10.1115/ICONE18-29737.

The present work investigates the turbulent jet flow mixing of downward impinging jets within a staggered rod bundle based on previous experimental work. The two inlet jets had Reynold’s numbers of 11,160 and 6,250 and were chosen to coincide with available data [Amini and Hassan 2009]. Steady state simulations were initially carried out on a semi-structured polyhedral mesh of roughly 13.2 million cells following a sensitivity study over six different discretized meshes. Very large eddy simulations were carried out over the most refined mesh and continuous 1D wavelet transforms were used to analyze the dominant instabilities and how they propagate through the system in an effort to provide some insight into potential problems relating to structural vibrations due to turbulent instabilities. The presence of strong standing horseshoe vorticies near the base of each cylinder adjacent to an inlet jet was noted and is of potential importance in the abrasion wear of the graphite support columns of the VHTR if sufficient wear particles are present in the gas flow.

Commentary by Dr. Valentin Fuster
2010;():693-700. doi:10.1115/ICONE18-29748.

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS.

Commentary by Dr. Valentin Fuster
2010;():701-704. doi:10.1115/ICONE18-29782.

The Single Event Effects (SEE) of Silicon On Insulator (SOI) and bulk-silicon NMOS are simulated using the SENTAURUS-TCAD device simulator. The Source-Drain Penetration Effect, which is caused by a heavy ion, was shown. It is proved that when the feature size of device become less than a certain scale, both Direct Channel Effect and Indirect Channel Effect occur. By comparing the distributions of equipotential lines in the MOSFETs’ channels of the different feature size devices during the ion strikes indirectly, the Source-Drain-Penetration Effect occurs more evidently when the device feature size is getting smaller.

Commentary by Dr. Valentin Fuster
2010;():705-712. doi:10.1115/ICONE18-29790.

SuperCritical Water-cooled nuclear Reactors (SCWRs) utilize a light-water coolant pressurized to 25 MPa with a channel inlet temperature of 350°C and outlet temperature of 625°C. Previous studies have indicated that uranium dioxide (UO2 ) nuclear fuel may not be suitable for SCWR use, because the maximum fuel centerline temperature might exceed the industry accepted limit of 1850°C. This research paper explores the use of uranium nitride (UN) as an alternative fuel option to UO2 at SuperCritical Water (SCW) conditions. A generic 1200-MWel Pressure-Tube (PT) -type reactor cooled with SCW was used for this thermalhydraulics analysis. The selected fuel option must have a fuel centerline temperature not higher than the industry accepted limit of 1850°C. Furthermore, the sheath (clad) temperature must not exceed the design limit of 850°C. The sheath and bundle geometry were adopted from previous studies. A single fuel channel was modeled using the UN fuel and an Inconel-600 sheath for several Axial Heat Flux Profiles (AHFPs). Uniform, upstream-skewed cosine, cosine and downstream-skewed cosine AHFPs were used. For each AHFP bulk-fluid, sheath and fuel centerline temperatures, and Heat Transfer Coefficient (HTC) profiles were calculated along the heated length of the channel. The calculations show that the UN fuel maintains a centerline temperature well below the industry accepted limit due to its high thermal conductivity at high temperatures. Therefore, the UN nuclear fuel is a viable fuel option for PT-type SCWRs.

Commentary by Dr. Valentin Fuster
2010;():713-719. doi:10.1115/ICONE18-29835.

The existence of boundary walls will impact flow and heat transfer to some extent. The boundary walls decrease the lateral velocities and change the flow direction to form some eddies, so the flow instability increases. Besides, they stop the mass transfer between the inner areas and outside. For the spacer grids of cluster fuel assembly, in order to confirm the scale of simulation object or experiment noumenon, the first thing is to research the influence depth of boundary effect, and try to eliminate its influence to the concerned object or area. This paper is focused on the spacer grids of 7×7 scale, 5×5 scale and 3×3 scale, respectively. Computed with CFD method, the lateral velocities of corner areas and typical subchannels are compared. And then, the influence depth of boundary effect is confirmed as the distance of a range of fuel rods. Therefore, for the concerned subchannel of CFD simulation, there must be no less than a range of fuel rods in the periphery to mitigate, so the result is not affected by the boundary effect.

Topics: Fuels , Manufacturing
Commentary by Dr. Valentin Fuster
2010;():721-727. doi:10.1115/ICONE18-29881.

We present advances in inverse transport methods and demonstrate their application to neutron tomography problems that have significant scattering. The problem we consider is inference of the material distribution in an object by detection and analysis of the radiation exiting from it. Our approach combines both deterministic and stochastic optimization methods to find a material distribution that minimizes the difference between computed and measured detector responses. The main advances are dimension-reduction schemes that we have designed to take advantage of known and postulated constraints. One key constraint is that the cross sections for a given region in the object must be the cross sections for a real material. We illustrate our approach using a neutron tomography model problem on which we impose reasonable constraints, similar to those that in practice would come from prior information or engineering judgment. This problem shows that our method is capable of generating results that are much better than those of deterministic minimization methods and dramatically more efficient than those of typical stochastic methods.

Commentary by Dr. Valentin Fuster
2010;():729-738. doi:10.1115/ICONE18-29918.

The reactor cavity cooling system (RCCS) for a very high temperature reactor (VHTR) represents a very important safety feature for achieving the defense in depth of the plant. An experimental facility was built for testing the heat transfer capability and phenomenology of this last heat sink designed for ensuring the cooling down of structural material of the vessel and of the concrete walls of the vessel cavity. This small scale facility was built using some of the scaling laws in order to resemble the main heat transport features in RCCS configuration. The natural convection phenomena and radiative heat transfer inside the cavity were represented. The experimental facility represents half of the vessel and of the reactor cavity with five stand pipes for cavity cooling using water as cooling fluid. Measurements were performed heating up the vessel surface temperature to an average temperature of 300 °C that is the average value in accident scenarios. Temperature measurements of the vessel surface temperature, the outer pipes surface temperature profile and inlet and outlet temperature of the cooling water were performed. Axial and radial temperature profiles of the air in the cavity were measured using a movable rack of 24 thermocouples. The results demonstrated the natural circulation phenomena. In addition Velocity measurement of the air inside the cavity were performed using particle tracking velocimetry techniques (PTV) determining the flow regime characteristics and the coupling with the temperature profile. The experimental test matrix of various flow rates in the cooling pipes were carried out.

Commentary by Dr. Valentin Fuster
2010;():739-745. doi:10.1115/ICONE18-29952.

This paper presents the analysis determining the status of fuel rods after whole normal operation. The FEMAXI–6 code was selected for such analysis. Evaluating the specifics of RBMK fuel rods, the adaptation of code was provided. After the adaptation of FEMAXI-6 code, the single fuel rod model of RBMK-1500 was developed and the processes, which occur during whole life of fuel rods, were analyzed. For this analysis the fuel rod from fuel channel with average initial power (2.5 MW) was selected. After (normal) operation the fuel rods from the reactor are transferred to the spent fuel pool and the state of the fuel rods (intactness of cladding, residual stresses in the cladding and fuel pellets, gap between cladding and pellets and etc.) is very important, because fuel rod cladding is one of the safety barriers. In this paper the stresses in cladding, plastic deformation of cladding and other parameters were calculated using FEMAXI-6 and method of final elements. The performed analysis demonstrates possibility to identify state of fuel rods after normal operation that is necessary for long-term fuel storage in spent fuel pools.

Topics: Modeling , Cycles , Fuel rods
Commentary by Dr. Valentin Fuster
2010;():747-754. doi:10.1115/ICONE18-29956.

Current CANDU-type nuclear reactors use a once-through fuel channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SCWR is to eliminate the annulus gap and use a ceramic insert to insulate the coolant from the moderator. An alternative approach being considered is to use the annulus gap as a coolant pre-heater in a double-pipe configuration (so-called re-entrant channel). The alternative design consists of two tubes, the inner tube (flow channel) and the outer tube (pressure tube). The fuel bundles, similar to those of the current CANDU reactors, are placed in the inner tube. The inner and outer tubes form an annulus through which flows the primary coolant. The coolant will flow through the annulus receiving heat from the inner tube from one end of the channel to another. At the far end, the flow will reverse direction and enter the inner tube, and hence the fuel bundle. At the inlet, the temperature is 350°C for a high-pressure coolant (pressure of 25 MPa), which is just below the pseudocritical point. At the outlet, the temperature is about 625°C at the same pressure (the pressure drop is small and can be neglected). Channel power and flow rate are variable initial conditions. The objective of this work is to model the heat transfer in the proposed fuel channel design. The channel has been divided into several nodes for the inner tube and the annulus gap. The power in each node in the pressure tube is considered to be uniform. The numerical model calculates the temperature profiles, the heat transfer coefficients and the overall heat transfer across the double-pipe fuel channel for a given set of flow, pressure and temperature boundary conditions and the power initial conditions. With the results from the numerical model, the design of the re-entrant (double-pipe) fuel channel can be optimized to improve its efficiency.

Commentary by Dr. Valentin Fuster
2010;():755-766. doi:10.1115/ICONE18-29972.

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 40-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30–35% to about 45–48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency Nuclear Steam Reheat (NSR) has to be introduced inside a reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat at subcritical-pressure experimental boiling reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the nuclear steam reheat at subcritical pressures in pressure-tube reactors is easier task than that in pressure-vessel reactors. Some design features related to the NSR are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors with the nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.

Commentary by Dr. Valentin Fuster
2010;():767-774. doi:10.1115/ICONE18-29974.

There are 6 prospective Generation-IV nuclear reactor conceptual designs. SuperCritical Water-cooled nuclear Reactors (SCWRs) are one of these design options. The reactor coolant in SCWRs will be light water operating at 25 MPa and up to 625°C, actually at conditions above the critical point of water (22.1 MPa and 374°C, respectively). Current Nuclear Power Plants (NPPs) around the world operate at sub-critical pressures and temperatures achieving thermal efficiencies within the range of 30–35%. One of the major advantages of SCWRs is increased thermal efficiency up to 45–50% by utilizing the elevated temperatures and pressures. SuperCritical Water (SCW) behaves as a single-phase fluid. This prevents the occurrence of “dryout” phenomena. Additionally, operating at SCW conditions allows for a direct cycle to be utilized, thus simplifying the steam-flow circuit. The components required for steam generation and drying can be eliminated. Also, SCWRs have the ability to support hydrogen co-generation through thermochemical cycles. There are two main types of SCWR concepts being investigated, Pressure-Vessel (PV) and Pressure-Tube (PT) or Pressure-Channel (PCh) reactors. The current study models a single fuel channel from a 1200-MWel generic PT-type reactor with a pressure of 25 MPa, an inlet temperature of 350°C and an outlet temperature of 625°C. Since, SCWRs are presently in the design phase there are many efforts in determining fuel and sheath combinations suited for SCWRs. The design criterion to determine feasible material combinations is restricted by the following constraints: 1) The industry accepted limit for fuel centreline temperature is 1850°C, and 2) sheath-material-temperature design limit is 850°C. The primary candidate fuel is uranium dioxide. However; previous studies have shown that the fuel centreline temperature of an UO2 pellet might exceed the industry accepted limit for the fuel centreline temperature. Therefore, investigation on alternative fuels with higher thermal conductivities is required to respect the fuel centreline temperature limit. Sheath (clad) materials must be able to withstand the aggressive SCW conditions. Ideal sheath properties are a high-corrosion resistance and high-temperature mechanical strength. Uranium dicarbide (UC2 ) is selected as a choice fuel, because of its high thermal conductivity compared to that of conventional nuclear fuels such as UO2 , Mixed OXide (MOX) and Thoria (ThO2 ). The chosen sheath material is Inconel-600. This Ni-based alloy has high-yield strength and maintains its integrity beyond the design limit of 850°C. This paper utilizes a generic SCWR fuel channel containing a continuous 43-element bundle string. The bulk-fluid, sheath and fuel-centreline temperature profiles together with Heat Transfer Coefficient (HTC) profile were calculated along the heated length of a fuel channel at the maximum Axial Heat Flux Profiles (AHFPs).

Commentary by Dr. Valentin Fuster
2010;():775-781. doi:10.1115/ICONE18-29975.

SuperCritical Water-Cooled nuclear Reactors (SCWRs) are one of six choices for Generation IV (Gen IV) reactor concepts. These reactors use light water as a coolant and operate at a pressure of 25 MPa, inlet temperatures 280–350°C and an outlet temperature up to 625°C. Operating at these elevated temperatures and pressures are beneficial due to: 1) increased gross thermal efficiency of SCW Nuclear Power Plants (NPPs) (from 30%–35% of the current NPPs to 45%–50%) and 2) decreased capital and operational costs. Use of SCW as a reactor coolant will permit a direct-cycle steam circuit. SCWRs eliminate the need for steam generators, steam separators, and steam dryers. Another advantage of SCWRs is a possibility for hydrogen co-generation through thermochemical cycles. At these extreme operating conditions we must be ensured that all fuel-channel materials, i.e., sheath (clad) and fuel, will operate below accepted temperature limits. The industry accepted limit for the fuel centerline temperature is 1850°C, and the design limit for sheath temperature is 850°C. Material investigations have begun with existing NPP fuel-channel designs. Previous studies with UO2 fuel at SCW conditions have indicated that the fuel centerline temperature may exceed the temperature limit. Zirconium alloys cannot operate at temperature beyond 350–500°C due to high corrosion rates. Therefore, Inconel-600 was chosen as a sheath material since is maintains a high yield strength and corrosion resistance at high temperatures. Uranium dioxide fuel is widely used and world resources are becoming limited. Thoria or thorium dioxide (ThO2 ) is considered as an alternative nuclear fuel and offers many benefits. Thorium dioxide is compliant to the Non-Proliferation Treaty, abundant in global reserves and has higher thermal conductivity than that of UO2 . An objective of this paper is to determine the suitability of ThO2 fuel in an Inconel-600-sheath fuel bundle within an SCWR fuel channel. Bulk-fluid, outer-sheath and fuel centerline temperature profiles along with Heat Transfer Coefficient (HTC) profiles were computed along the heated length of a bundle string at the maximum heat flux.

Topics: Water , Nuclear fuels
Commentary by Dr. Valentin Fuster
2010;():783-792. doi:10.1115/ICONE18-29990.

This paper presents an extensive study of heat-transfer correlations applicable to supercritical-water flow in vertical bare tubes. A comprehensive dataset was collected from 33 papers by 27 authors, including more than 125 graphs and wide ranges of parameters. The parameters ranges were as follows: pressures 22.5–34.5 MPa, inlet temperatures 85–350°C, mass fluxes 250–3400 kg/m2 s, heat fluxes 75–5,400 kW/m2 ), tube heated lengths 0.6–27.4 m, and tube inside diameters 2–36 mm. This combined dataset was then investigated and analyzed. Heat Transfer Coefficients (HTCs) and wall temperatures were calculated using various existing correlations and compared to the corresponding experimental results. Three correlations were used in this comparison: Bishop et al., Mokry et al. and modified Swenson et al. The main objective of this study was to select the best supercritical-water bare-tube correlation for HTC calculations in: 1) fuel bundles of SuperCritical Water-cooled Reactors (SCWRs) as a preliminary and conservative approach; 2) heat exchangers in case of indirect-cycle SCW Nuclear Power Plants (NPPs); and 3) heat exchangers in case of hydrogen co-generation at SCW NPPs from SCW side. From the beginning, all these three correlations were compared to the Kirillov et al. vertical bare-tube dataset. However, this dataset has a limited range of operating conditions in terms of a pressure (only one pressure value of 24 MPa) and one inside diameter (only 10 mm). Therefore, these correlations were compared with other datasets, which have a much wider range of operating conditions. The comparison showed that in most cases, the Bishop et al. correlation deviates significantly from the experimental data within the pseudocritical region and actually, underestimates the temperature at most times. On the other hand, the Mokry et al. and modified Swenson et al. correlations showed a relatively better fit within the most operating conditions. In general, the modified Swenson et al. correlation showed slightly better fit with the experimental data than other two correlations.

Commentary by Dr. Valentin Fuster
2010;():793-799. doi:10.1115/ICONE18-29991.

Many heat-transfer correlations exist for bare tubes cooled with SuperCritical Water (SCW). However, there is very few correlations that describe SCW heat transfer in bundles. Due to the lack of extensive data on bundles, a limited dataset on heat transfer in a SCW-cooled bundle was studied and analyzed using existing bare-tube correlations to find the best-fit correlation. This dataset was obtained by Razumovskiy et al. (National Technical University of Ukraine “KPI”) in SCW flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of tubes of 5.2-mm outside diameter and 485-mm heated length. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within a range from 800 to 3000 kg/m2 s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2 . The objective of this study is to compare bare-tube SCW heat-transfer correlations with the data on 1- and 3-rod bundles. This work is in support of SuperCritical Water-cooled Reactors (SCWRs) as one of the six concepts of Generation-IV nuclear systems. SCWRs will operate at pressures of ∼25MPa and inlet temperatures of 350°C.

Topics: Heat transfer
Commentary by Dr. Valentin Fuster
2010;():801-808. doi:10.1115/ICONE18-29998.

Generation IV nuclear reactor technology is increasing in popularity worldwide. One of the six Generation-IV-reactor types are SuperCritical Water-cooled Reactors (SCWRs). The main objective of SCWRs is to increase substantially thermal efficiency of Nuclear Power Plants (NPPs) and thus, to reduce electricity costs. This reactor type is developed from concepts of both Light Water Reactors (LWRs) and supercritical fossil-fired steam generators. The SCWR is similar to a LWR, but operates at a higher pressure and temperature. The coolant used in a SCWR is light water, which has supercritical pressures and temperatures during operation. Typical light water operating parameters for SCWRs are a pressure of 25 MPa, an inlet temperature of 280–350°C, and an outlet temperature up to 625°C. Currently, NPPs have thermal efficiency about of 30–35%, whereas SCW NPPs will operate with thermal efficiencies of 45–50%. Furthermore, since SCWRs have significantly higher water parameters than current water-cooled reactors, they are able to support co-generation of hydrogen. Studies conducted on fuel-channel options for SCWRs have shown that using uranium dioxide (UO2 ) as a fuel at supercritical-water conditions might be questionable. The industry accepted limit for the fuel centerline temperature is 1850°C and using UO2 would exceed this limit at certain conditions. Because of this problem, there have been other fuel options considered with a higher thermal conductivity. A generic 43-element bundle for an SCWR, using uranium mononitride (UN) as the fuel, is discussed in this paper. The material for the sheath is Inconel-600, because it has a high resistance to corrosion and can adhere to the maximum sheath-temperature design limit of 850°C. For the purpose of this paper, the bundle will be analyzed at its maximum heat flux. This will verify if the fuel centerline temperature does not exceed 1850°C and that the sheath temperature remains below the limit of 850°C.

Topics: Fuels , Uranium , Water , Heat flux
Commentary by Dr. Valentin Fuster
2010;():809-817. doi:10.1115/ICONE18-30024.

This paper presents an analysis of heat transfer in water at supercritical conditions in bare vertical tubes. A large dataset within conditions similar to those of SuperCritical Water-cooled nuclear Reactors (SCWRs) was obtained from the Institute for Physics and Power Engineering (Obninsk, Russia). This dataset was compared to existing heat-transfer correlations from the open literature. This comparison is an extension to the previous studies done with this dataset. Previous studies have shown that existing correlations, such as the Dittus-Boelter correlation significantly overestimates the experimental heat transfer coefficient (HTC) values within the pseudocritical range; the Bishop et al. and Jackson’s correlations were also found to deviate significantly from the experimental data. The Swenson et al. correlation provided a better fit for the experimental data, as compared to the previous three correlations within some flow conditions, but deviates from data for other conditions. HTC and wall temperature values calculated with the FLUENT CFD code also deviate from the experimental data within some conditions. After analyzing the existing correlations, it was decided to develop a better correlation for predicting HTC. Since the Swenson et al. correlation seems to be the best candidate for predicting the experimental data; it was selected as a basis for developing a new empirical correlation. The primary difference of the Swenson et al. approach is that it uses the majority of thermophysical properties at the wall temperature as opposed to those used at bulk-fluid temperatures in other models. Calculating various thermophysical properties based on wall temperature seems to give much better results in terms of accuracy. To obtain a basic empirical correlation, a dimensional analysis was conducted using a combination of various dimensionless terms. This approach was combined with the experimental dataset at the normal heat-transfer regime using statistical analysis. The final correlation showed the best fit for the experimental dataset within a wide range of flow conditions. The calculated wall temperatures were within (±15%) for the analyzed dataset, which is a considerable improvement from the previous correlations. The accuracy of calculated values was further improved when a term was added to the correlation that accounted for the entrance effect in bare tubes. Thus, the new correlation presented in this paper can be used for HTC calculations in supercritical-water heat exchangers at SCW Nuclear Power Plants (NPPs) in case of the indirect cycle, in heat exchangers for co-generation of hydrogen from supercritical water side, for a preliminary heat-transfer calculations in SCWR fuel channels as a conservative approach. It can also be used for future comparisons with other independent datasets, with bundled data, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids.

Commentary by Dr. Valentin Fuster
2010;():819-824. doi:10.1115/ICONE18-30063.

The CANDLE (C onstant A xial shape of N eutron flux, nuclide densities and power shape D uring L ife of E nergy production) burnup strategy is a new burnup concept. The CANDLE reactors generate energy by using only natural or depleted uranium as make up fuel and achieve about 40% burnup without fuel recycling of the conventional nuclear energy concept. So far the CANDLE cores feature a relatively large peak-to-average power density and discharge burnup distribution. Peaked power and burnup distribution are undesirable as they deteriorate economical performance. The objective of this paper is to study the feasibility of power flattening of sodium cooled large scale CANDLE reactor toward commercial use by using thorium fuel loading into the inner core zone. When power density profile becomes flat, it is expected that the axial position of burning region is aligned at the same height for each radial position. It makes core height shorter and raises the average power density farther. The shorter core has usually more merits such as smaller loss of coolant pressure obtained during passing fuel channel and more negative coolant void coefficient. For this purpose, thorium is added uniformly to the uranium fuel in the inner core. If we choose the amount of thorium proper, net radial current of neutrons in the inner core becomes zero in the inner core, and at the boundary between inner and outer core enough neutrons leak from the uranium region and the net radial current is still zero at this point. In the outer region the neutrons leak outward. By this way, we can make the power density distribution flat in the inner core. In the present work, the power density profile is intended flatten for the metallic fuel CANDLE reactors by adding thorium uniformly in the inner core region. The maximum axially integrated power density (radial peaking factor) decreases from 1.87 with only uranium fuel to 1.44 with uranium and thorium fuels. We can expect increasing average discharge burnup and decreasing fuel inventory and pressure drop.

Topics: Combustion , Fuels , Design , Sodium
Commentary by Dr. Valentin Fuster
2010;():825-833. doi:10.1115/ICONE18-30069.

The objective of this paper is to calculate heat losses from a CANDU-6 fuel-channel while modifying it according to the specified operating pressure and temperature conditions of SuperCritical Water-cooled Reactors (SCWRs). Heat losses from the coolant to the moderator are significant in a SCWR because of high operating temperatures (i.e., 350–625°C). This has adverse effects on the overall thermal efficiency of the Nuclear Power Plant (NPP), so it is necessary to determine the amount of heat losses from fuel-channels proposed for SCWRs. Inconel-718 was chosen as a pressure tube (PT) material and PT minimum required thickness was calculated in accordance with the coolant’s maximum operating pressure and temperature. The heat losses from the fuel-channel were calculated along the heated length of the fuel-channel. Steady-state one-dimensional heat-transfer analysis was conducted, and programming in MATLAB was performed. The fuel-channel was divided into small segments and for each segment thermal resistances of the fuel-channel components were analyzed. Further, the thermophysical properties of the coolant, annulus gas, and moderator were retrieved from the NIST REFPROP software. The analysis outcome resulted in a total heat loss of 29.3 kW per fuel-channel when the pressure of the annulus gas was 0.3 MPa.

Commentary by Dr. Valentin Fuster
2010;():835-843. doi:10.1115/ICONE18-30073.

Gas-liquid two-phase flows are widely encountered in industrial plants such as chemical reactors, power plants, environmental plants and so on. It is essential to clarify the dissolution process of the bubbles and the structure of the gas-liquid two-phase flows for realizing the efficient operation of these industrial reactors. However, it is very difficult to clarify them because of their complexity. We made a challenge of clarifying them in detail. First, we discuss the dynamical processes of the mass transfer from a zigzagging CO2 bubble of 2.9 mm in equivalent diameter by using three measurement methods (i.e. LIF (laser induced fluorescence)/HPTS (8-hydroxypyrene-1, 3, 6-trislfonic acid), PIV and a newly developed photoelectric optical fiber probe: POFP) effectively and mutually-complementarily. We directly visualized the dynamical mass transfer process from a zigzagging CO2 bubble to the surrounding liquid by using LIF/HPTS. We measured the surrounding liquid motion induced by the bubble buoyancy using PIV. We visualized a high-CO2 -concentration thin layer around the bubble was transported to the bubble rear and accumulated into the horseshoe-like vortices. The clear horse-shoe-like vortices were observed just after the launch of the bubble from a needle. Then, the wreckage of the bubble wake was transported widely into the surrounding liquid by the buoyancy driven flows. We measured directly the CO2 concentration profile inside the bubble wake by using POFP. We succeeded in clarifying the profile of the CO2 concentration inside the bubble wake, which is difficult to obtain from only the LIF/HPTS method. From this result, we obtained that the CO2 concentration takes the maximum at the center region of the bubble wake and sharply decreases toward the outer edge of the bubble wake. Second, we simultaneously measured the diameters and velocities of the bubble by using POFP. We confirmed the performance of the POFP; the results via POFP were compared with those obtained from the visualization of the bubbles by using a high-speed video camera. We demonstrated mutually-complementary use of three measurement methods is very effective to experimentally understand the dynamical processes of the mass transfer from a zigzagging CO2 bubble to the surrounding liquid.

Commentary by Dr. Valentin Fuster
2010;():845-854. doi:10.1115/ICONE18-30104.

At present, there are a number of Generation-IV nuclear reactor concepts under development worldwide, and the SuperCritical Water-cooled nuclear Reactor (SCWR) type is one of them. The main objective of developing SCWRs is to: 1) Increase the thermal efficiency of current Nuclear Power Plants (NPPs) from 30–35% to approximately 45–50%, and 2) Decrease capital and operational costs. SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25 MPa and outlet temperatures up to 625°C). This paper presents a SCWR single-reheat indirect cycle concept with intermediate heat exchangers. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, heat exchangers separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in the reactor building. The nuclear activities stay within the reactor building, and there is a reduced possibility for radioactive contamination of equipment in the turbine building. As SCW NPPs will have much higher operating thermal hydraulic parameters this paper analyzes the technical challenges and higher costs typically associated with heat exchangers. The double-pipe heat exchanger is analyzed in depth to determine the heat-transfer surface area, number of units and physical dimensions of the heat exchanger. This study will help to determine whether the advantages of the indirect cycle justify implementation of heat exchangers at a SCW NPP.

Commentary by Dr. Valentin Fuster
2010;():855-864. doi:10.1115/ICONE18-30116.

A defense-in-depth feature for advanced light water reactors to cope with beyond design basis accidents is the ability to cool and stabilize ex-vessel core melt debris. Several international experimental programs have investigated core-concrete interactions and debris cooling of ex-vessel core melts. These experimental programs have identified various phenomena which affect melt coolability and may enhance it. One such phenomenon, melt eruptions, occurs when gas from the underlying decomposing concrete entrains melt up through and onto a solidified crust, which separates the molten melt from the cooling water. Previous modeling and experimental work have shown this cooling mechanism can have a large impact on melt coolability. Previous melt eruption models are reviewed and a new synthesis model is proposed. Reviewing past experimental evidence and modeling efforts indicate the geometry of the flow area impacts the amount of melt ejected. To understand the potential flow area available for melt eruptions, past experimental evidence is reviewed, a steady state analysis of flow area is performed and non-steady state considerations are discussed.

Topics: Concretes
Commentary by Dr. Valentin Fuster
2010;():865-871. doi:10.1115/ICONE18-30119.

The loss of air conditioning in the electrical auxiliary building at South Texas Project (STP) has been shown in previous work to result in relatively rapid air temperature rise. The heat up is a concern in the Probabilistic Risk Assessment (PRA) because it may be associated with high conditional core damage probability. As a consequence, operator responses to mitigate the building heat up have been developed and proceduralized at STP. The plant’s current loss of electrical auxiliary building HVAC operator procedure was analyzed and improvements have been recommended. This was done using a commercial computational fluid dynamics (CFD) package that modeled the transient air temperature rise of the building and qualitatively assessed the effectiveness of operator actions. A possible methodology for validating the CFD results was developed using coupled energy balance equations for individual rooms. The methodology is a conservative approach to compare the effectiveness operator actions and provide steps for future higher fidelity simulation and validation. While the specific approach is applicable to STP, the overall methods and approaches described should be applicable to other sites that may be subject to excessive room heat up due to loss of air conditioning in critical rooms.

Topics: Temperature
Commentary by Dr. Valentin Fuster
2010;():873-878. doi:10.1115/ICONE18-30148.

The experimental study has been carried out to investigate the behavior of ejection and sweep events in a rectangular channel turbulent flow. The channel is a 1800 mm long vertical channel with a cross sectional area of 100 mm and 20 mm. The events are studied by obtaining velocity components of the flow by using the ultrasonic measurement method (UVP). This paper clarifies the capability of the UVP for investigating the sweep and ejection events.

Commentary by Dr. Valentin Fuster
2010;():879-885. doi:10.1115/ICONE18-30153.

Several new fuel assembly designs for multi-recycling Transuranics from spent nuclear fuel are proposed and investigated. Among these are (1) Mixed Oxide Fuel with Enriched Uranium (MOX-EU), in which Plutonium oxide and U-235 enriched Uranium oxide are mixed (2) MOX fuel with Americium coating, in which a thin layer of Americium is applied to the outer surface of the MOX fuel pellet, and (3) an heterogeneous fuel assembly consists of Inert-Matrix Fuel (IMF) pins at the periphery and UOX pins in the inner zone. All these designs are compatible with standard PWR utilizing 17×17 fuel assemblies. In-reactor fuel depletion simulation and long-term isotopic decay calculation are carried out using DRAGON[1] and ORIGEN[2], separately. Transuranics mass balance and long-term radiotoxicity analyses are implemented and the results are normalized to per 1TWh-electricity produced.

Commentary by Dr. Valentin Fuster
2010;():887-893. doi:10.1115/ICONE18-30173.

It is necessary to consider the complexities of both natural and engineered components of a nuclear waste repository since fission products and minor actinides remain harmful to the environment for tens of thousands of years. In safety and performance assessments often used in decision-making about repository designs, the effect of uncertain initial guesses on the models’ output must be understood. As the necessary safe times and hence the simulated times are often in the order of magnitude of hundreds of thousands of years, uncertain initial values become increasingly important. To minimize the danger from high-level radioactive waste and to make informed decisions over designs, sensitivity analysis of the models used should be performed. The Simplified Total System Performance Assessment (STSPA) model developed by Golder Associates Inc., Booz-Allen Hamilton, Stone and Webster and the University of Nevada Reno and used in the Yucca Mountain nuclear waste repository performance assessment is analyzed for sensitivity by varying the activities of technetium-99 and iodine-129 by several orders of magnitude. The resultant dose to a maximally-exposed individual over time periods of 100,000 and 1,000,000 years is compared to the relevant regulatory limits. Incorrect estimates can be seen to have large effects on the behavior of the model while the method used allows conclusions to be drawn about the robustness of the model.

Commentary by Dr. Valentin Fuster
2010;():895-900. doi:10.1115/ICONE18-30177.

Reactor power and neutron activity control is the main key for safe reactor operation. Reactivity coefficients and effects are main measures to estimate reactor control and safety. These characteristics outline reactors behavior during usually exploitation and accident events. Reactivity coefficients and effects quantify the effect, which various parameters (e.g. fuel and graphite temperatures, amount of steam) have for the core neutron activity. Many modifications of RBMK-1500 reactor cores in Ignalina NPP were made during their lifetime. Reactor core modifications like load of higher enriched fuel with burnable absorber and new design control rods affected reactivity coefficients and effects. Neutron-physical parameters calculations of reactor core states with variant fuel loads and new design control rods were performed using QUABOC/CUBBOC-HYCA software. The changes of reactivity coefficients and effects were quantified in this paper.

Commentary by Dr. Valentin Fuster
2010;():901-912. doi:10.1115/ICONE18-30192.

In support of developing SuperCritical Water-cooled Reactors (SCWRs), studies are currently being conducted for heat-transfer at supercritical conditions. This paper presents an analysis of heat-transfer to SuperCritical Water (SCW) flowing in bare vertical tubes as a first step towards thermohydraulic calculations in a fuel-channel. A large set of experimental data, obtained in Russia, was analyzed. Two updated heat-transfer correlations for forced convective heat transfer in the normal heat transfer regime to SCW flowing in a bare vertical tube were developed. It is expected that the next generation of water-cooled nuclear reactors will operate at supercritical pressures (∼25 MPa) with high coolant temperatures (350–625°C). Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare tube data can be used as a conservative approach. The analyzed experimental dataset was obtained for SCW flowing upward in a 4-m-long vertical bare tube. The data was collected at pressures of about 24 MPa for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values for mass flux ranged from 200–1500 kg/m2 s, for heat flux up to 1250 kW/m2 and inlet temperatures from 320–350°C. The Mokry et al. correlation was developed as a Dittus-Boelter-type correlation, with thermophysical properties taken at bulk-fluid temperatures. Alternatively, the Gupta et al. correlation was developed based on the Swenson et al. approach, where the majority of thermophysical properties are taken at the wall temperature. An analysis of the two updated heat-transfer correlations is presented in this paper. Both correlations demonstrated a good fit (±25% for Heat Transfer Coefficient (HTC) values and ±15% for calculated wall temperatures) for the analyzed dataset. Thus, these correlations can be used for preliminary HTC calculations in SCWR fuel bundles as a conservative approach, for SCW heat exchangers, for future comparisons with other independent datasets and for the verification of computer codes for SCWR core thermohydraulics.

Topics: Heat transfer , Water
Commentary by Dr. Valentin Fuster
2010;():913-922. doi:10.1115/ICONE18-30225.

Rod bundles are widely used in industry today with applications ranging from nuclear reactors, heat exchangers, and steam generators. Accurately modeling the inherently unsteady and turbulent flow within these rods is essential in order to design for optimal efficiency while controlling vibration, noise and heat transfer. The problem complicates further when spacer-grids are used within the rods to maintain separation and structural rigidity. Computational modeling can be a useful alternative to the costly process of manufacturing and testing prototypes, but its accuracy needs to be checked with detailed experimental data. This paper describes an experimental database obtained using two-dimensional Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 × 5 rod bundle with spacer-grids. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction technique employed in this investigation which consists of immersing plastic rods with a similar index of refraction as the one for water to achieve optical transparency across the spacer grid. This unique feature allows flow visualization and measurement within the bundle without rod obstruction. This approach also allows the use of high temporal and spatial non-intrusive dynamic measurement techniques namely TR-PIV to investigate the flow evolution below and immediately above the spacer. The data base presented includes explanation of the various cases tested such as boundary conditions, rig dimensions, measurement zones, and the test equipment in order to provide a good base for Computational Fluid Dynamics (CFD) simulations. Turbulence analysis of the obtained data is provided in order to gain insight of the physical phenomena and to compare the possible results obtained from computational simulations.

Topics: Fuels , Turbulence
Commentary by Dr. Valentin Fuster
2010;():923-928. doi:10.1115/ICONE18-30347.

This paper presents a numerical study of the particle behaviors under acceleration conditions in the solid-air two-phase flow by means of a combined two-dimensional model of computational fluid dynamics and discrete element method (CFD-DEM). The simulation model provides the information regarding the particle distribution behaviors within the calculation region and the particle run-out rate from the calculation region under different parameter conditions, such as particle size, initial particle loading and particle acceleration condition. The results demonstrate that the particle run-out rate is directly influenced by the particle size and the initial loading condition. The particle acceleration in the horizontal direction adversely affects the particle run-out rate when the initial particle loading condition is dispersed and uniform. However, this adverse effect disappears when the initial particle loading condition becomes concentrate and partial.

Commentary by Dr. Valentin Fuster

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