neutron transport

The necessity of keeping in working condition the sources of traditional electricity generation for dispatching the supply of electricity from the capacities of "green" electricity generation has been realized. Potentially, one of these sources may be nuclear power plants. An obstacle to their use as a participant in dispatching is the impossibility of maneuvering with power in a wide range. One of the reasons for limiting the range is the complexity of automatic compensation for the reactivity of the xenon oscillation reactor. To

The necessity of keeping in working condition the sources of traditional electricity generation for dispatching the supply of electricity from the capacities of "green" electricity generation has been realized. Potentially, one of these sources may be nuclear power plants. An obstacle to their use as a participant in dispatching is the impossibility of maneuvering with power in a wide range. One of the reasons for limiting the range is the complexity of automatic compensation for the reactivity of the xenon oscillation reactor. Existing physical and mathematical models for calculating the parameters of processes in the reactor due to changes in its power because of its complexity cannot be used in operational automatic control systems. The task of constructing an approximation linear model of processes in the reactor in the form of a transfer function is set. To build an approximation model, the inverse problem is solved. The desired model is based on the condition of coin...

The water-cooled ceramic breeder (WCCB) blanket is one candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), and it is being developed at the Institute of Plasma Physics of the Chinese Academy of Sciences. This paper presents an updated WCCB blanket concept design that is expected to cover the CFETR operation modes of 200 MW, 500 MW, 1 GW, and 1.5 GW fusion power and achieve tritium self-sufficiency with the latest core parameters (major radius R = 7.2 m; minor radius a = 2.2 m) and mission. The feasibility of the new blanket design is evaluated considering neutronics and thermal-hydraulics aspects.

This thesis evaluates the nodal synthesis method for nuclear reactor transient analysis. The basic idea of the nodal synthesis method is to expand the time-dependent neutron fluxes as a linear combination of a set of pre-computed static trial functions with unknown timedependent mixing coefficients. The mixing coefficients are then obtained from a weightedresidual approach. The main purpose is to investigate the feasibility of the nodal synthesis model for real-or near real-time application to tightly coupled cores such as the MIT Research Reactor. The approach consists of two portions. First, the theoretical nodal synthesis model, which is based on the coarse-mesh, finite-difference neutron diffusion equations with discontinuity factors, is studied. The neutron fluxes are expanded and substituted into those governing equations, and then the mixing coefficients are calculated by the weighted-residual method. Several one-dimensional, steady-state and transient test problems are analyzed to evaluate the accuracy and computational efficiency of the theoretical nodal synthesis model. The second part is the validation of an approach which can be taken to determine the few-group nodal parameters for the application of the experimental nodal synthesis method to transient studies involving the MIT Research Reactor. The approach uses the Monte-Carlo code MCNP to generate few-group nodal cross sections and discontinuity factors for the triangular-z nodal diffusion code QUARTZ. A simple test problem was analyzed to demonstrate the procedure. The main findings are that (1) the nodal synthesis model can reproduce reference neutron-flux distributions with an average error in total reactor power of about 1-2 %; (2) the computing speed for the nodal synthesis model is not fast enough for practical applications; and (3) good statistical behavior and the determination of accurate groupto-group scattering cross sections are two key factors for the complete validation of the MCNP-QUARTZ procedure.

TRIPOLI® and TRIPOLI-4® are registered trademarks of CEA.International audienceAlthough the Monte Carlo (MC) codes are natural users of the fast growing capacities in High PerformanceComputing (HPC), adapting production level codes such as TRIPOLI-4 to the hexascale is very challenging. Wepresent here the dual strategy we follow new thoughts and developments for the next versions of TRIPOLI-4, aswell as insights on a prototype of next generation Monte Carlo (NMC) designed from the beginning with hexascalein mind.The Random Generators of the code will also be presented as well as the strategy of verificationof the parallelism

A new prototype display tool T4G for TRIPOLI-4 Monte Carlo transport code is being developed and validated. This paper explains the basics of the tool and demonstrates its objectives -easy to use, interactive with user, fast display, great help to build and debug geometry and material input data, and useful to check variance reduction parameters and mesh tally output. It improves the final quality of TRIPOLI-4 input by better error checking and reduces the input files maintenance cost for industrial and engineering projects. For the TRIPOLI-4 team, T4G is also a useful tool to speed up considerably the discovering of unknown geometries generated by different users manually and/or by computer programs. The recently introduced CAD conversion interface programs for ITER neutronics and LMJ building shielding calculations generate huge TRIPOLI-4 inputs with more than twenty thousands volume cells and fifty materials. The T4G tool was proven to be very helpful to fast navigate and to quick debug the CAD conversed TRIPOLI-4 inputs, to study the variance reduction techniques, and to check the colorful image results from the mesh tally simulation output.

In this updated overview (cf. [1]) of the Monte Carlo transport code TRIPOLI-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. TRIPOLI-4 has support for execution in parallel mode. Special features and applications are also presented.

The accelerator-driven system (ADS) is an innovative reactor that is being considered as a dedicated high-level-waste burner in a double-strata fuel cycle. ("Double-strata fuel cycle" means a partitioning and transmutation system for long-lived radioactive nuclides.) The target is the physical and functional interface between the accelerator and the subcritical reactor in the ADS, so it is probably the most innovative component of the ADS. Key parameters of ADS are the number of neutrons emitted per incident proton, the neutron multiplicity (n/p), the mean energy deposited in the target for neutrons produced, the neutron energy spectrum, and the spallation product spatial distribution. This paper focuses on the production of neutrons in the spallation reactions. The neutrons produced in the spallation reactions can be characterized by their energy and spatial distributions and multiplicity. The present calculations have been performed using the Monte Carlo code MCNPX. The Monte Carlo simulations have been performed to investigate the neutron multiplicity as a function of incident proton beam energy, as well as a function of target material and target size. Neutron flux distributions at the target surface are calculated and compared with different target materials and proton energies. A comparison of MCNPX with experimental results is made.

In the fuel management of nuclear reactors, which has a direct impact on power generation costs and environmental factors, it is essential to obtain an optimal arrangement of assemblies and core parameters. This complex task relies on both neutronics and thermal hydraulics. This research solves this by employing PARCS and COBRA-EN codes, alongside Multi-Verse Optimization algorithm (MVO) as a novel algorithm. Findings demonstrate the MVO algorithm has reliability and favorable convergence rate. Moreover, integrating the nodal expansion technique with MVO notably decreases the computational time for optimization in the loading pattern optimizations (LPOs). The optimized loading pattern aligns well with the findings in the literature. Also, using multi objective fitness function leads to improving core flatness, cycle length and safety parameters.

One of the generic designs of the nuclear fusion DEMO reactor proposed by the EUROfusion consortium foresees the development of a tritium Breeding Blanket (BB) relying on the use of the liquid-metal PbLi eutectic alloy as both neutron multiplier and tritium breeder, namely the Water-Cooled Lithium Lead (WCLL) BB, whose strengths and weaknesses are well known. This paper focuses the attention on one of the possible disadvantages of such a technology: the production of the highly radiotoxic radionuclide 210Po, which could become a safety issue to be accounted for. The 210Po concentration within the PbLi circuit has been assessed by solving a modified version of Bateman’s equations to consider the alloy circulation, so a one-dimensional convective fluid-dynamic model has been set up. Nuclear quantities have been evaluated by Monte Carlo neutron transport analyses using MCNP code and adopting a fully heterogeneous model of DEMO equipped with the WCLL BB. Moreover, rough sensitivity anal...

The shielding analysis of the spent fuel dry storage cask TN-32 is carried out using the continuous-energy Monte Carlo neutron-and photon-transport code MCS developed by the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The shielding analysis involves the transport simulation of the decay neutrons and gammas emitted by the spent fuel to obtain the neutron and gamma dose rate values outside of the cask. This deep-penetration problem is solved by means of weight window population control as variance reduction. The TN-32 cask calculations are repeated with the MCNP6 Monte Carlo code for verification purpose and a good agreement within three standard deviations is observed between MCS and MCNP6. The neutron dose rate on the cask surface and 2 m away from the cask can be calculated with MCS, whereas the photon dose rate calculation for positions outside of the cask will be further improved by employing more sophisticated variance reduction techniques.

This paper deals with spectral theory of a new class of neutron transport operators involving collision operators of the form K = K i + K e where K i (resp. K e ) describes the inelastic (resp. elastic) collisions of neutrons with the host material. We give a fine analysis of their asymptotic point spectra for isotropic space homogeneous cross sections in bounded geometries. We provide a new formalism relying on spectral analysis of some non compact symmetrizable operators arising in transport theory. Additional results on essential spectra are also given.

In this review, we detail the commonality of mathematical intuitions that underlie three numerical methods used for the quantitative description of electron swarms propagating in a gas under the effect of externally applied electric and/or magnetic fields. These methods can be linked to the integral transport equation, following a common thread much better known in the theory of neutron transport than in the theory of electron transport. First, we discuss the exact solution of the electron transport problem using Monte Carlo (MC) simulations. In reality we will progress much further, showing the interpretative role that the diagrams used in quantum theory and quantum field theory can play in the development of MC. Then, we present two methods, the Monte Carlo Flux and the Propagator method, which have been developed at this moment. The first one is based on a modified MC method, while the second shows the advantage of explicitly applying the mathematical idea of propagator to the transport problem.

In the course of discussing different target types for their suitability in the European Spallation Source (ESS) one main focus was on neutronics' performance. Diverse concepts have been assessed baselining some preliminary engineering and geometrical details and including some optimization. With the restrictions and resulting uncertainty imposed by the lack of detailed designs optimizations at the time of compiling this paper, the conclusion drawn is basically that there is a little difference in the neutronic yield of the investigated targets. Other criteria like safety, environmental compatibility, reliability and cost will thus dominate the choice of an ESS target.

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Neutronics calculations have been performed of the MYRRHA ADS Reactor with a thorium-based fuel mixture, using the simulation programs MCNPX [1] and Geant4 . Thorium is often considered for ADS systems, and this is the first evaluation of the possibilities for thorium based fuels using a reactor design which has been developed in detail. It also extends the application of the widely-used Geant4 program to the geometry of MYRRHA and to thorium. An asymptotic 232 Th/ 233 U mixture is considered, together with the standard MOX fuel and a possible 232 Th/MOX starter. Neutron fluxes and spectra are calculated at several regions in the core: fuel cells, IPS cells and the two (Mo and Ac) isotope production cells. These are then used for simple calculations of the fuel evolution and of the potential for the incineration of minor actinide waste. Results from the two programs agree and support each other and show that the thorium fuel is viable, and has good evolution/breeding properties, and that minor actinide incineration, though it will not take place on a significant scale, will be demonstrable.

Almract--Ditticulties can arise in radiative and neutron transport calculations when a highly anisotropic scattering phase function is present. In the presence of anisotropy, currently used numerical solutions are based on the integro-differential form of the linearizecl Boltzmann transport equation. In this paper, we depart from classical thought and present an alternative numerical approach based on application of the integral form of the transport equation. Use of the integral formalism facilitates the following steps: a reduction in dimensionality of the system prior to diseretization, the use of symbolic manipulation to augment the computational procedure, and the direct determination of key physical quantities which are derivable through the various Legendre moments of the intensity. Our approach is developed in the context of radiative heat transfer in a plane-parallel geometry, and results are presented and compared with existing benchmark solutions. Encouraging results are presented to illustrate the potential of the integral formalism for computation. The integral formalism appears to possess several computational attributes which are well-suited to radiative and neutron transport calculations.

The singular-eigenfunction-expansion method and the principle of invariance are combined to reduce the two-half-space Milne problem to a regular computational form in the two-group isotropic scattering model. The method used here consists in considering a problem of two contiguous half-spaces with surface sources at the interface. The problem is equivalent to the Milne problem in the sense that the expansion coefficients are to be determined from the same equation. The emergent distributions are obtained from coupled regular integral equations. The expansion coefficients can then be obtained using the halfrange orthogonality relation of the eigenfunctions. Numerical results are reported for light-water media.

The singular-eigenfunction-expansion method and the principle of invariance are combined to reduce the two-half-space Milne problem to a regular computational form in the two-group isotropic scattering model. The method used here consists in considering a problem of two contiguous half-spaces with surface sources at the interface. The problem is equivalent to the Milne problem in the sense that the expansion coefficients are to be determined from the same equation. The emergent distributions are obtained from coupled regular integral equations. The expansion coefficients can then be obtained using the halfrange orthogonality relation of the eigenfunctions. Numerical results are reported for light-water media.

The IRT-2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an ...

The effects of a strongly anisotropic scattering law on the critical slab problem in monoenergetic neutron transport theory are studied in this paper using a synthetic kernel. The Boltzmann equation is converted into a single integral equation, which is solved rigorously by a simple and fast convergent projection procedure. Numerical results for critical parameter and critical flux are presented and compared to those already obtained by Siewert and Williams via different semianalytical algorithms.

O Hipotireoidismo, definido como alteracao generalizada dos processos metabolicos, advindos da diminuicao da funcao da tireoide, e consequente reducao ou ausencia da producao de seus hormonios, pode ser acompanhado de serias sequelas no desenvolvimento, razao pela qual se justificam os programas de triagem neonatal. Ja esta estabelecida uma relacao direta entre hipotireoidismo congenito e as deficiencias auditivas perifericas, entretanto, pouco sabe- se sobre as alteracoes no sistema nervoso auditivo central, na presenca da disfuncao tireoideana. Destaca-se que o hormonio tireoideano e essencial para o desenvolvimento cerebral normal, sendo um importante regulador das sinapses, dos processos de mielinizacao, sintese e acao dos neurotransmissores. Desta forma, objetivou-se estudar as vias auditivas centrais, atraves dos potenciais evocados auditivos de curta e media latencia em criancas diagnosticadascom Hipotireoidismo Congenito, em tratamento hormonal, triadas pelo Programa Estadua...

Yellow color denotes the unresolved resonance region, gray color denotes the resolved resonance region, and MC 2 -2R represents the MC 2 -2 calculation with RABANL option. * Yellow color denotes the unresolved resonance region, gray color denotes the resolved resonance region, and MC 2 -2R represents the MC 2 -2 calculation with RABANL option. * Yellow color denotes the unresolved resonance region, gray color denotes the resolved resonance region, and MC 2 -2R represents the MC 2 -2 calculation with RABANL option. * Yellow color denotes the unresolved resonance region, gray color denotes the resolved resonance region, and MC 2 -2R represents the MC 2 -2 calculation with RABANL option. * Yellow color denotes the unresolved resonance region, gray color denotes the resolved resonance region, and MC 2 -2R represents the MC 2 -2 calculation with RABANL option. * Yellow color denotes the unresolved resonance region, gray color denotes the resolved resonance region, and MC 2 -2R represents the MC 2 -2 calculation with RABANL option. * Yellow color denotes the unresolved resonance region, gray color denotes the resolved resonance region, and MC 2 -2R represents the MC 2 -2 calculation with RABANL option.

This paper presents the new computational capabi lity of MC2-3 for generation of multigroup cross section in thermal energy range. The multi-gr oup cross section generation code MC -3 for fast reactor applications has recently been extended to generate the cross sections for the entire energy range of interest in fission reactors. Thermal scattering matrices an d i teraction cross section libraries for the energ y ange from 10-5 eV to 5.0 eV were prepared with the NJOY code in a 1700 group structure. The slowing-down and transport calculation capabilities of MC 2-3 were extended to 10 -5 eV in compliance with the new thermal group cross section libraries. Numerical tests were fo LWR fuel pin and assembly problems. The 3478-group neutron spectra obtained with MC 2-3 agreed well with MCNP6 results. The eigenvalue e rror was less 160 pcm and the pin power error was less than 1.2% comp ared to MCNP6 results. These results indicate that t e modified MC2-3 can generate accurate cross sect...

This paper presents the neutronics modeling capabilities of the fast reactor simulation system SHARP, which ANL is developing as part of the U.S. DOE's NEAMS program. We discuss the three transport solvers (PN2ND, SN2ND, and MOCFE) implemented in the UNIC code along with the multigroup cross section generation code MC 2 -3. We describe the solution methods and modeling capabilities, and discuss the improvement needs for each solver, focusing on massively parallel computation. We present the performance test results against various benchmark problems and ZPR-6 and ZPPR critical experiments. We also discuss weak and strong scalability results for the SN2ND solver on the ZPR-6 critical assembly benchmarks.

This report presents the effort under way at Argonne National Laboratory toward a comprehensive, integrated computational tool intended mainly for the high-fidelity simulation of sodium-cooled fast reactors. The main activities carried out involved neutronics, thermal hydraulics, coupling strategies, software architecture, and highperformance computing. A new neutronics code, UNIC, is being developed. The first phase involves the application of a spherical harmonics method to a general, unstructured three-dimensional mesh. The method also has been interfaced with a method of characteristics. The spherical harmonics equations were implemented in a stand-alone code that was then used to solve several benchmark problems. For thermal hydraulics, a computational fluid dynamics code called Nek5000, developed in the Mathematics and Computer Science Division for coupled hydrodynamics and heat transfer, has been applied to a single-pin, periodic cell in the wire-wrap geometry typical of advanced burner reactors. Numerical strategies for multiphysics coupling have been considered and higher-accuracy efficient methods proposed to finely simulate coupled neutronic/thermalhydraulic reactor transients. Initial steps have been taken in order to couple UNIC and Nek5000, and simplified problems have been defined and solved for testing. Furthermore, we have begun developing a lightweight computational framework, based in part on carefully selected open source tools, to nonobtrusively and efficiently integrate the individual physics modules into a unified simulation tool Results reported in the AFCI series of technical memoranda frequently are preliminary and subject to revision. Consequently, they should not be quoted or referenced without the authors' permission.

If important phenomena cannot be calculated by analysis tools, then further development is undertaken. Analysis: The operational and accident scenarios that require study are analyzed No Yes Yes Figure E-1. PIRT informed R&D process. Form 412.09 (Rev. 10

A design for the demonstration HTGR has not yet been selected. Consequently, the R&D process is focused on scenarios and highly ranked phenomena that have already been identified as important by the advanced gas-cooled reactor community for all the designs being considered as candidates for the HTGR. This approach has resulted in an HTGR-specific Phenomenon Identification and Ranking Table (PIRT) from which the methods R&D is being defined using the following assumptions: 1. The selected NGNP design could be either a pebble-bed or a prismatic reactor. a. Reasonable agreement is achieved when the calculation generally lies within the uncertainty band of the data used for validation and always shows the same trends as the data. Code deficiencies are minor. Form 412.09 (Rev. 10

Monte Carlo criticality calculations can be subject to slow fission source convergence for large or weakly coupled problems. By tallying the spatially-discretized fission kernel (the linear operator stochastically applied during power iteration), a low-order fundamental eigenvector is calculated to accelerate fission source convergence. In MCNP the fission matrix is stored efficiently using a sparse structure, the tallying of which requires little extra work. In addition, higher eigenmodes and eigenvalues from the fission matrix are calculated, which conventional Monte Carlo is unable to produce. After testing this method on several realistic problems, significant convergence acceleration was found, along with higher eigenmode information.

The neutronics analysis of the current core of the TRIGA Mark II research reactor is performed at the Atominstitute (ATI) of Vienna University of Technology. The current core is a completely mixed core having three different types of fuels i.e. aluminium clad 20% enriched, stainless steel clad 20% enriched and SS clad 70% enriched (FLIP) Fuel Elements (FE(s)). The completely mixed nature and complicated irradiation history of the core makes the reactor physics calculations challenging. This PhD neutronics research is performed by employing the combination of two best and well practiced reactor simulation tools i.e. MCNP (general Monte Carlo N-particle transport code) for static analysis and ORIGEN2 (Oak Ridge Isotop Generation and depletion code) for dynamic analysis of the reactor core. The PhD work is started to develop a MCNP model of the first core configuration (March 1962) employing fresh fuel composition. The neutrons reaction data libraries ENDF/B-VI is applied taking the missing isotope of Samarium from JEFF3.1. The MCNP model of the very first core has been confirmed by three different local experiments performed on the first core configuration. These experiments include the first criticality, reactivity distribution and the neutron flux density distribution experiment. The first criticality experiment verifies the MCNP model that core achieves its criticality on addition of the 57th FE with a reactivity difference of about 9.3 cents. The measured reactivity worths of four FE(s) and a graphite element are taken from the log book and compared with MCNP simulated results. The percent difference between calculations and measurements ranges from 4 to 22%. The neutron flux density mapping experiment confirms the model

The purpose of this study is to develop a complete Monte Carlo model of Chashma Nuclear Power Plant Unit-2 (CNPP-II) core and benchmark against the theoretical (Final Safety Analysis Report FSAR) and experimental results. For this purpose, a comprehensive model of CNPP-II core is developed using Monte Carlo N-Particle (MCNP5) radiation transport code which is equipped with temperature dependent ENDF/B-VII.1 cross-section library. The model is validated against the reference results of CNPP-II at seven different operating conditions of the initial core. For Cold Zero Power (CZP) condition, the overall temperature is taken as 40 C, while for Hot Zero Power (HZP) condition, the temperature is taken as 280 C. In Hot Full Power (HFP), the fuel, cladding and coolant temperatures are taken as 873 C, 340 C and 302 C respectively. To enhance the model validation from global level (k eff ) to local level (flux or power distribution), the simulated pin & assembly wise neutron flux and power distribution is compared with the available experimental results of CNPP-II. The comparison demonstrates a reasonable agreement between computational and reference results. This study infers that the criticality value and excess reactivity are within design limits of CNPP-II and the reactor has a symmetrical flat power distribution, ensuring the safe reactor operation. The deviations between the calculated and experimental results have been explained in this paper. This model is interesting in the sense that it presents a new full core benchmark with experimental validation. This research also reveals that CNPP-II, which is in SMR category, is a safe and reliable reactor.

The Atominstitute (ATI) of Vienna University of Technology (VUT) operates a TRIGA Mark II research reactor since March 1962. Its initial criticality was achieved on 7th March 1962 when 57th Fuel Element (FE) was loaded to the core. This paper describes the development of the MCNP model of the TRIGA reactor and its validation through three different experiments i.e. initial criticality, reactivity distribution and a thermal flux mapping experiment in the reactor core. All these experiments were performed on the initial core configuration. The MCNP model includes all necessary core components i.e. FE, Graphite Element GE, neutron Source Element (SE), Central IRradiation channel (CIR) etc. Outside the core, this model simulates the annular grooved graphite reflector, the thermal and thermalizing column, four beam tubes and the reactor water tank up to 100 cm in radial and +60 and -60 cm in axial direction. Each grid position at its exact location is modeled. This model employs the ENDF/B-VI data library except for the Sm-isotopes which are taken from JEFF 3.1 because ENDF/B-VI lacks samarium (Sm) cross sections. For the first experiment, the model predicts an effective multiplication factor (Ä eff ) of 1.00183 with an estimated standard deviation 0.00031 which is very close to the experimental value 1.00114. The second experiment measures the reactivity values of four FE and one GE. In comparison to the MCNP results, the percent difference ranges from 4 to 22. The third experiment verifies the model at a local level with the radial and axial thermal flux density distribution in the core. Though the trends are similar, the MCNP model overestimates the radial thermal flux density in the core and underestimates these results at the core periphery.

Gamma spectrometry is one of the common methods to inspect the spent fuel from research reactors. This method has been applied to in-pool measurements of the Spent Fuel Elements (SPEs) of the TRIGA Mark II research reactor. Due to mixed nature of the reactor core and complicated irradiation history of the fuel elements (FEs), the gamma spectrometry of the FE establishes improvements in the calculation and measurement of the SPE. In order to inspect the TRIGA SPE from dry storage and cooled fuel from the reactor pool, the selected spend fuels are scanned and measured using the fuel-scanning machine. Gamma spectrometry is performed by HPGe detector for spend fuel inspection and determination of the 137 Cs activity and 134 Cs/ 137 Cs ratio. In this work, the steps of the detector calibration and the use of the Monte Carlo radiation transport code (MCNP5) have been described. In addition, the fuel-scanning machine and the gamma spectrometer are modelled by MCNP5 to simulate the gamma transport from fuel to detector. It also simulate the gamma spectrometer calibration for the burn up determination of the spend fuel. The results from MCNP5 simulation are applied to spectroscopic measurements and compared with the theoretical predictions of the neutronics code ORIGEN2 in this research work.

In this work, the formulation of the Analytical Discrete Ordinates (ADO) method for two-dimensional fixed-source neutron transport problems is extended to problems with anisotropic scattering. The methodology is developed from the discrete ordinates approximation of the two-dimensional transport, whose equations are integrated transversally within homogenized regions of the domain, yielding onedimensional equations for the average angular fluxes. Such one-dimensional equations are solved by the ADO method, after considering approximations for the transverse leakage terms, and their solutions are written in terms of eigenvalues and eigenfunctions. The relevant feature of the ADO method for the isotropic scattering cases, which reduced the order of the eigenvalue problems to half of the number of discrete directions, is also preserved for the present anisotropic cases. Explicit spatial variables expressions are derived for the average angular fluxes in regions of interest defined in the domain. The full domain solution is defined by the coupling of local solutions in different regions, without the use of sweeping procedures. As usual in nodal schemes, auxiliary equations are necessary and, in this case, the unknown region-edge angular fluxes are approximated by constants embedded in the source term. Numerical results for region-average scalar fluxes are obtained and compared with the ones available in the literature to illustrate the feasibility of keeping the computational efficiency already verified in treating problems with isotropic scattering, with the use of the ADO method.

A neutronics mock-up of the ITER inboard shielding system including first wall, shield-blanket, vacuum vessel and magnetic field coils, was investigated in order to validate calculational tools used for the ITER design and to verify the nuclear performance of the shield by measuring neutron and photon flux spectra at two positions inside the bulk assembly. The measured spectra are compared with results from calculations using the Monte Carlo code MCNP-4A and the data library FENDL-1. Integral values such as neutron-induced reaction rates and gamma-heating are calculated on the basis of the experimentally obtained flux spectra, and are compared with independent measurements of these data. Helium and hydrogen production rates, as well as the radiation damage rate, are determined again by using the measured flux spectra.

A neutronics mock-up of the International Thermonuclear Experimental Reactor (ITER) inboard shielding system with streaming path for the fusion neutrons was investigated in order to validate the design. Neutron and photon flux spectra were measured at several positions in the assembly. The experimentally determined flux spectra are analysed with the ITER design tools Monte Carlo code MCNP-4A and Fusion Evaluated Nuclear Data Library FENDL-1 and FENDL-2. Helium production, radiation damage and photon heating rates, which are design parameters of the shielding system are derived from both, the measured and the calculated spectral fluxes.

Reactor design requires safety studies to ensure that the reactors will behave appropriately under incidental or accidental situations. Safety studies often involve multiphysics simulations where several branches of reactor physics are necessary to model a given phenomenon. In those situations, it has been observed that the neutron transport part is still a bottleneck in terms of computational times, with more than 80% of the total time. In the case of hexagonal lattice reactors, transport solvers usually invert the discretised Boltzmann equation by discretising the regular hexagon into lozenges or triangles. In this work, we seek to reduce the computational burden of the neutron transport solver by designing a numerical spatial discretisation scheme that would be more appropriate for honeycomb meshes. In our past research efforts, we have set up interesting discretisation schemes in the finite element setting in 2D, and we wish to extend them to 3D geometries that are prisms with a hexagonal base. In 3D, a rigorous method was derived to shrink the tensor product between 2D and 1D bases to minimum terms. We have applied these functions successfully on a reactor benchmark-Takeda Model 4-to compare and contrast the numerical results in a physical setting.

, a German ophthalmologist (Photo 10.1), postulated intuitively the relationship between a chemical concentration flux and its gradient in a solution by analogy with Fourier's law in heat conduction. It was only 50 years later that Albert Einstein demonstrated the accuracy of Fick's law through his work on Brownian motion. In neutron physics, a neutron scatters through matter with successive collisions that reduce its energy to thermal energy, whence the analogy with collisions of non-interacting molecules in a gas. The free scattering path is the sum of the elementary distances covered by the neutron between each collision (with atoms that constitute matter) from its birth till its loss through absorption (length of the dotted line in Fig. .1). The mean free scattering path λ s is the mean distance normalized to one collision. It is obtained by dividing the free scattering path by the number of collisions. Fick's law linearly relates a physical quantity to its derivative. It formulates the idea that a quantity diffuses in the direction of its gradient in the same way that salt in salted water diffuses from zones of higher concentration to zones of lower concentration or that heat transfers from hotter zones to cooler ones. Applying this relation to neutron physics, the relation between the neutron current and its flux is similar to that which exists in the heat equation between the temperature field and

Industrial reactors have complex geometries, with the nuclear fuel being geometrically separated from the coolant. Furthermore, different materials are used to allow for mechanical and thermal stresses due to the emitted energy. Hence, the homogeneous model cannot be applied in all situations, which leads to more complex neutron models for calculations. (Meghreblian and Holmes 1960, p. 626) At this point, it is clear that a homogeneous reactor is the simplest case to calculate. However, it is not exactly representative of either neutron physics or the technology employed in energy production. During its lifetime in the reactor, a neutron is slowed down, reaching epithermal energy where several resonances are present, especially those of 238 92 U, which are abundantly present in the reactor. Slowing-down in the resonance zone is a critical phase during which almost 30% of neutrons do not "survive" (the resonance escape probability is about 0.7). If a homogeneous reactor is considered with a homogeneous solution of fissile nuclides and moderator, the continuous slowing-down process by collision with the moderator increases the chances of the neutron being captured in resonances. Separating the moderator from the fuel increases the probabilities that neutrons emitted from the fuel then slowed down by the moderator will escape resonant capture (Marchuk 1959, p. 167). The reference work, (Kahan and Gauzit 1957), illustrates this effect by showing the impact of the ratio of concentrations of graphite to natural uranium in moles on the k 1 of a homogeneous solution (Fig. ). An optimum moderation ratio of N graphite /N Uranium ¼ 400 is obtained, i.e. k max 1 ¼ 0:8. Already in 1942, Enrico Fermi was conscious that for a maximum fp of just 0.6, the infinite multiplication factor k 1 of a homogeneous reactor with natural

SIMMER-III, a neutronics and thermal-hydraulics coupled code, was originally developed for core disruptive accident analyses of liquid metal cooled fast reactors. Due to its versatility in investigating scenarios of core disruption, the code has also been extended to the simulation of transients in thermal neutron systems such as the criticality accident at the JCO fuel fabrication plant, and, in recent years, applied to water-moderated thermal research reactor transient studies, too. Originally, SIMMER considered only cylindrical fuel pin geometry. Therefore, implementation of a plate-type fuel model to the SIMMER-III code is of importance for the analysis of research reactors adopting this kind of fuel. Furthermore, validation of the SIMMER-III modeling of light water-cooled thermal reactor reactivity initiated transients is of necessity. This paper presents the work carried out on the SIMMER-III code in the framework of a KIT and IRSN joint activity aimed at providing the code with experimental reactor transient study capabilities. The first step of the job was the implementation of a new fuel model in SIMMER-III. Verification on this new model indicates that it can well simulate the steady-state temperature profile in the fuel. Secondly, three cases with the shortest reactor periods of 5.0 ms, 4.6 ms and 3.2 ms among the Special Power Excursion Reactor Tests (SPERT) performed in the SPERT I D-12/25 facility have been simulated. Comparison of the results between the SIMMER-III simulation and the reported SPERT results indicates that although there is space for further improvement on the modeling of negative feedback mechanisms, with this plate-type fuel model SIMMER-III can well represent the transient phenomena of reactivity driven power excursion.

An axial fuel shuffling strategy is proposed based on the mechanism of the nuclear fission traveling wave and implemented numerically in the calculation for a supercritical water cooled fast reactor (SCWFR). The ERANOS code is adopted to perform the neutronics and burn-up calculations, and the calculation scheme for axial fuel shuffling and coolant density coupling are set up. The parametric studies of a typical PWR with TheU and UePu ( 235 U instead of 239 Pu) conversions by burn-up and k eff calculations indicate that the breeding effects only exist in configurations with very low water content and the conversion or breeding becomes worse as the initial enrichment is increasing. The shuffling calculations for the 1-D SCWFR model described in this paper brought about some interesting results for a certain range of water content. The results indicate that the non-enriched fresh fuel is not possible for both TheU and U ePu conversions. As could be expected due to the h-values of the main fissile isotopes 233 U and ( 235 U, 239 Pu), respectively, the TheU conversion needs a lower enrichment, and results in a slightly higher burn-up than the UePu conversion. The asymptotic power density distribution of the TheU conversion is broader and lower than that of the UePu conversion. By reducing the water volume fraction, an increased burn-up can be achieved with correspondingly reduced fuel shuffling speed and reduced initial enrichment. Furthermore, the steady state calculations for the asymptotic state show that the TheU conversion is superior to the UePu one concerning SCWFR safety aspects, where the absolute value of the Doppler constant is larger and the coolant feedback is negative for the TheU conversion, while the coolant feedback is positive for the UePu one.