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Highlights • Potential advantages of 99Mo production in an Aqueous Homogeneous Reactor (AHR) are discussed. • An AHR conceptual design using low-enriched uranium for the production of 99Mo is studied. • Aspects related to the neutronic behavior of the AHR were studied using MCNPX code. • The reactivity feedback introduced by the production of radiolytic gas bubbles was analyzed. • The feasibility of 99Mo production in an Aqueous Homogeneous Reactor is discussed. Abstract Nowadays, 99mTc is the most common radioisotope used in nuclear medicine, with up to 30–40 million procedures worldwide every year. Furthermore, medical diagnostic imaging techniques using 99mTc represent approximately 80% of all nuclear medicine procedures. Currently, 99mTc is almost exclusively produced from the beta-decay of its 66-h parent 99Mo. The 99Mo production in an Aqueous Homogeneous Reactor (AHR) is potentially advantageous because of its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This paper studies an AHR conceptual design using low-enriched uranium for the production of 99Mo. Aspects related to the neutronic behavior such as critical height, medical isotopes production, uranium consumption, plutonium production and the reactivity feedback introduced in the solution by the volumetric expansion of the fuel solution were evaluated using the computational code MCNPX version 2.6e. In addition, important reactor kinetic parameters such as the effective delayed neutron fraction, β eff , and mean neutron generation time, Λ , were calculated. A benchmarking exercise was solved using available results of critical experiments performed at the Russian Research Center "Kurchatov Institute". The neutronic calculations demonstrated that the reactor is able to produce 99 six-day Ci of 99Mo in operation cycles of five days. The reactivity feedbacks introduced by the volumetric expansion of the fuel solution is at about −1460 pcm, which represents approximately the 48% of the planned initial reactivity reserve in the core. The calculated effective delayed neutron fraction and the average mean neutron generation time were 785 pcm and 134.08 µs, respectively. [ABSTRACT FROM AUTHOR]

Highlights • Assessment of detectors' locations and number of detectors in PGSFR. • Generation of spatial weighting function using forward fixed-source calculation in MCNPX. • Calculation of transient power during withdrawal of a single control assembly by using FREK. Abstract This paper presents the assessment of the ex-core detectors in the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) in response to the movement of single primary control assembly, to evaluate the detectors' locations and to determine a proper number of detectors. The detectors are located below the containment vessel and distributed in different ranges from the center and different azimuthal angles. Another position considered is in the radial direction of the core, at the sodium coolant gap between the outermost assembly and core shroud. To calculate the detection rate, the contribution of each fuel region to the detector and the core power distribution are required. Initially, the aforementioned contribution is generated by performing a continuous-energy forward fixed-source Monte Carlo calculation using MCNPX. The reactor kinetics code FREK simulates the power change due to an inadvertent withdrawal of a primary control assembly up to 10% from its critical position followed by full insertion of secondary control assemblies. The response from each detector location was evaluated, and consequently, the detectors located below the containment vessel, which can "see" all the fuel assemblies, responded better to the core power increase with a minor error of less than ∼0.6%. In addition, the impact of locating the detector below the containment vessel at different locations from the center or at different azimuthal angles was insignificant but several detectors should be distributed at even azimuthal angles for effective core monitoring and redundancy purposes. [ABSTRACT FROM AUTHOR]

For long times there has been a sustained international interest in partitioning and transmutation of spent fuel produced by fission reactors, in particular both plutonium (Pu) and minor actinides (MA), which are associated with nuclear proliferation risks and radioactivity hazards. Transmutation requires the Pu and MA to be irradiated in a very intense neutron flux that can only be attained in a high power fission system such as a thermal reactor, a fast reactor or an accelerator driven system (ADS). The present work investigated two proposed models of core design of different Pu and MA distribution (uniform and nonuniform) applied to TRIGA Mark II ADS reactor to maximize fuel utilization, covering energy extraction and Pu/MA mass reduction. Transmutation performance was studied through burnup and mass ratio time evolution calculations of fuel and minor actinides various components including U, Pu, Am and Cm isotopes, up to two years operation cycle using MCNPX Transport code. [ABSTRACT FROM AUTHOR]

The broad goal of this work is to use the accumulated plutonium around the world in peaceful uses of energy production rather than for military purposes. This work investigates the feasibility of burning the Pu vector that is extracted from the spent fuel of light water reactor or from the nuclear warheads stockpile. MCNPX code version 2.7 has been used to design two assembly models, conventional assembly and Blanket-seed (BS) assembly. The conventional assembly is fueled with UO 2 while the BS assembly is fueled with ThO 2 in the blanket region and Pu vector either in the form of reactor grade plutonium (rgPu) or weapon grade plutonium (wgPu) in the seed region. Using of Th-232 as a fertile material in the blanket region prevents from more Pu breeding and enables a longer fuel burnup rate. Five different fuel models are studied to investigate the optimum concentration of Pu vector that is used in the seed region. MCNPX code is used to simulate the neutronic characteristics of the presented fuels. The multiplication factor K eff values of the suggested fuel in the BS assembly have been analyzed and compared with that obtained in the conventional assembly. Some parameters related to safety operation such as moderator temperature coefficient (MTC) and effective delayed neutron fraction (β eff ) have been calculated. The Pu vector concentrations have been investigated with burnup to distinguish the effectively of the suggested models to burn how much of the used Pu in the seed region. Using of Th-232 as a fertile material in the blanket region enhance the ability of BS assembly to bred more fissile material (U-233). Therefore, the suggested fuel assemblies achieve a higher conversion ratio (CR). The radioactivity of the actinides and non-actinides elements has been investigated with burnup. Due to the poising effect of fission products on the reactivity of the reaction, it is important to investigate the concentrations of the most important fission product at the presented fuel. [ABSTRACT FROM AUTHOR]

In this paper thorium fuel feasibility in large scale Gas Cooled Fast Reactor (GCFR) is investigated. The neutronics benchmark used in this study, GFR2400, corresponds to a 2400 MWth GFR concept proposed by the French CEA. MCNPX computational code is used to design a 3D heterogeneous model of the GFR2400 core. A detailed feasibility analysis of the performance of thorium fuel cycle is performed by using thorium as an alternative fertile fuel for the natural uranium vector of the reference core design. The most essential neutronic parameters characterizing the core are determined both for beginning of life (BOL) conditions as well as during burnup. Also, a three-dimensional core cycle-by-cycle simulations is performed to allow explicit characterization of the core behavior and safety-related parameters during both open and equilibrium cycles. The thorium-based core shows favorable neutronic characteristics with an acceptable control and safety parameters for both BOL and open cycle states. The depressurization reactivity effect and core expansion coefficients (axial and radial) show improvement compared to the uranium-based core. However, this improvement is compensated by the deterioration in the effective delayed neutron fraction (β-eff) and the Doppler reactivity Effect. The results of isotopic transmutation and fuel burnup confirm the capability of the core to work in both open and closed cycles and to self-recycle its own MA vector and plutonium of LWRs. However-as is the case in other fast reactors including the uranium-based GFR2400 core, the fuel cycle closure causes safety related parameters to degrade. [ABSTRACT FROM AUTHOR]

The pursuit for an optimum burnable absorbers (BAs) takes a great interest due to their importance for regulating and control the reactivity of the reactor. MCNPX code is used to simulate the effect of Europium and Pyrex on the neutronic characteristics of PWR and compare it with the conventional burnable absorber. The thermal neutron flux and the power distribution through PWR assembly are analyzed to investigate the effect of gadolinium and europium on it. The PWR assembly is filled in a different fuel burnup process with 2% Gd 2 O 3 and 2% Eu 2 O 3 as integral burnable absorbers (IBAs). Despite the fact that the absorption cross section of gadolinium is larger than europium, the effective multiplication factor in the case of (UO 2 + Gd 2 O 3 ) IBAs rods increases more than that in the case of (UO 2 + Gd 2 O 3 ) IBAs rods. The Pyrex alloy is investigated in the Burnable poison rods (BPRs) to compare its effect on the neutronic characteristics of PWR with the conventional material such as B 4 C and (Ag-In-Cd). The effect of these materials on the Keff values has been analyzed. Due to the high absorption cross section of 135 Xe and 149 Sm resulting from the fission process, it was important to study the effect of different BAs materials on their concentration. The 239 Pu and 241 Pu concentrations resulting from the burning process have been calculated for the suggested BAs materials. [ABSTRACT FROM AUTHOR]

Fast reactor concept like the 2400 MW GCFR is considered a promising candidate for future sustainable and economic nuclear power systems. It is a highly innovative system with advanced geometry and fuel materials (Ceramic fuel pellets of mixed uranium-plutonium carbide within fuel pins). The active fuel region of this reactor is divided radial to two core zones namely, inner core fuel assemblies (IC), and outer core fuel assemblies (OC). In the present study Monte Carlo transport code, MCNPX is used to design heterogeneous and homogeneous models for inner and outer core fuel assembly. The deigned models have been used to simulate the neutronic behavior and fuel transmutation during open cycle operation and to study the effect of Homogeneity inside the fuel assembly. The results of heterogeneous and homogeneous models were compared with each other and good agreement was found. [ABSTRACT FROM AUTHOR]

• Neutron radiography. • MCNPX. • IECF. • Mont-Carlo Simulation. One of the most useful non-destructive techniques for imaging low Z materials hidden behind high Z materials is fast neutron radiography (FNR). Inertial Electrostatic Confinement Fusion (IECF) as a small neutron source is an excellent apparatus for neutron radiography because of its ability to generate high-flux fast neutrons. In this study, the effects of geometric parameters on images obtained by a FNR system were investigated using MCNPX2.7 Monte-Carlo code. Geometric unsharpness can be reduced by optimizing the source-object and object-detector distance. The results showed that the contribution of scattered neutrons is approximately 0.1 % of that of transmitted ones for object-detector distances greater than five times the size of the side of the object perpendicular to the beam axis. The optimal detector-object distance was 50 cm. By increasing the source- object distance, better contrast and a higher fitting between the image and the object in terms of the position was observed. The optimal source-object distance was considered 150 cm. [ABSTRACT FROM AUTHOR]

• MCP alloys including Zn, Ga, Sn, Pb and Bi elements was studied as shielding. • Photon energies of 0.01 to 10 MeV were considered. • The MCNPX code, XCOM, XMuDat and Auto-Z eff programs were used. • The µ m , HVL, σ a , Z eff and N eff values were calculated for MCP samples. • The calculated data were compared with available experimental data. Photon attenuation properties of low melting-point MCP alloys of 5 N Plus Company are evaluated for 10 keV–10 MeV using MCNPX code, XCOM and XMuDat programs, Auto-Z eff software and linear interpolation method. The simulated and calculated results were in good agreement with differences of < ± 2 %. MCP-124 had the highest shielding efficiency compared to other alloys. It was found that, the µ m and σ a values of studied alloys decreased with photon energy. MCP-200 had the highest values of 2.7 HVL at the 6 MeV photon energy. The σ a values of MCP alloys increased as follow at the photon energies above 0.1 MeV: MCP-69 < MCP-200 < MCP-137 < MCP-96 < MCP-124. In addition, the Z eff values averaged on studied photon energy range were 30.86, 47.74, 68.14, 76.63 and 82.55 respectively for MCP-69, MCP-200, MCP-137, MCP-96 and MCP-124 alloys. The N eff values of alloys varied in the range of 1.63 × 1023 – 3.03 × 1023 electron/g. [ABSTRACT FROM AUTHOR]

Plutonium (Pu) and minor actinides (MAs) are responsible for the majority of radiotoxicity, heat generation, and proliferation risk of nuclear spent fuel. Accelerator driven system (ADS) reactors can employ Pu and MA for transmutation and power generation within improved safety scheme. Pu and MA can be loaded in ADS reactor core in a variety of distributions. In this study, the effects of uniform and nonuniform distributions of Pu and MA samples loaded within enriched uranium fuel elements inside UT-TRIGA Mark II ADS reactor on neutron flux, neutron spectrum, fission rate, and subcritical multiplication parameters were numerically investigated using Monte Carlo MCNPX transport code. The type of distribution producing higher subcritical multiplication and source efficiency will contribute in minimizing the accelerator current requirements for ADS operation. [ABSTRACT FROM AUTHOR]

A cold neutron source (CNS) conceptual designing for the Tehran Research Reactor (TRR) were carried out using MCNPX code. In this study, a horizontal beam tube of the core which has appropriate the highest thermal flux is selected and parametric analysis to choose the type and geometry of the moderator, and the required CNS dimensions for maximizing the cold neutron production was performed. In this design the moderator cell has a spherical annulus structure, and the cold neutron flux and its brightness are calculated together with the nuclear heat load of the CNS for a variety of materials including liquid hydrogen, liquid deuterium, and solid methane. Based on our study, liquid hydrogen with more ortho-concentration than para and solid methane are the best options. [ABSTRACT FROM AUTHOR]

The existence of significant local intra-assembly and azimuthal neutron flux variations at or around PWR control rod fingers and which would eventually not be captured by lower order transport deterministic reactor analysis methods, was examined using simplified full-core Monte Carlo calculations. The locations of specific interest were the clad surfaces and absorber material zones around the tips of partially inserted control rods in controlled assemblies located at various core locations. A fixed-source option was adopted in order to reduce the modeling efforts and computational costs. However, the validity of the fixed-source approach for predictions of the local flux distributions was verified by comparisons against corresponding calculations performed in eigenmode and against deterministic CASMO/SIMULATE solutions. Noticeable azimuthal neutron flux variations up to ∼30% were observed but only for the thermal energy range. For the fast (>1 MeV) neutron flux, the predicted azimuthal variations were found to be bounded by an upper value of about ∼10%. On the other hand, significant planar flux non-uniformities between individual control rods within a controlled assembly were seen for all energies. Hence, in addition of having allowed for a strengthened knowledge on the behavior of local flux variations around PWR control rods, these results also provide a first assessment of the likely uncertainty level in control rod tip fluence estimations based on lower resolution deterministic reactor analysis codes. [ABSTRACT FROM AUTHOR]

The accelerator driven system (ADS) is a very interesting option to improve the safety of nuclear power reactor and for transmutation of spent fuel. The Texas phase of the reactor–accelerator coupling experiment (RACE), completed in March 2006, demonstrated the feasibility of operating a training research isotopes general atomic (TRIGA) research reactor in a subcritical configuration driven to a significant power by an electron LINAC neutron source (photoneutron). In the present study, the effects of changing the source cylindrical target material, radius, position and the electron beam energy on the final neutron production, fission probability, and the subcritical system multiplication of TRIGA Mark II research reactor, have been numerically investigated. Three target materials are used: Tungsten, Lead and Tungsten–Copper alloy, while varying the target radius from 2 to 8 cm, the source position at three locations, and the beam energy from 10 to 55 MeV. The investigation is based on the numerical calculation of the subcritical multiplication factor and the external source efficiency using Monte Carlo MCNPX code. Through the comparison of the studied cases results, the favorable target material and radius, source position, and beam energy can be obtained. [ABSTRACT FROM AUTHOR]

The main target of this research was to estimate the possibility for improving the performance of Boiling Water Reactor (BWR) cores by using uranium thorium dioxide fuels instead of the commonly used oxide fuel. MCNPX (Monte Carlo N-Particle eXtend) computer code based on Mont Carlo method, is used to design two models for a BWR fuel bundle in a typical operating temperature and pressure conditions, the first is the oxide fuel bundle model and the second is the blanket–seed (BS) bundle model. A test case was compared with a previously published data calculated by HELIOS code and good agreement was found. The two models are used to determine the multiplication factor, pin-by-pin power distribution and thermal neutron flux distribution for both UO 2 and (UZr/ThO 2 ) fuels. BS bundle model improves the performance of BWR in different ways: (a) by increasing the energy extracted per fuel assembly, (b) by reducing the uranium ore, (c) reducing the plutonium accumulated in BWR through burn-up and continues to grow throughout the world, and (d) by increasing fuel burn-up and reactor lifetime. The spent fuel in the reactor can be recycled, and the plutonium and its isotopes can be extracted. The extracted plutonium can be used as a Mixed Oxide (MOX) fuel. [ABSTRACT FROM AUTHOR]

This paper presents a design of boiling water reactor BWR model using MCNPX to develop benchmarks for checking the fuel management computer code packages. MCNPX code based on Monte Carlo method, is used to design a three dimensional model for BWR fuel assembly in typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. This design is used to study the thermal neutron flux and the pin by pin power distribution through the BWR core assemblies. The fuel used in BWR core is UO 2 with three different types of enrichment (0.711%, 1.76% and 2.19%). This enrichment is distributed in such a way as to flatten the power. The effect of different enrichment values on the radial normalized power distribution is analyzed. The spent fuel in the reactor can be recycled, and plutonium and its isotopes can be extracted. [ABSTRACT FROM AUTHOR]

Stochastic sampling (SS) method for quantifying nuclear data uncertainties is accomplished by using perturbed nuclear data in routine neutronics calculations and determining the variance of output parameters due to the input nuclear data uncertainties. Existing SS-based methods have demonstrated the feasibility and efficiency of propagating uncertainties in multigroup nuclear data. However, in fields such as criticality safety assessment, pointwise representation of nuclear data is more appropriate in order to corroborate the increasing safety demand and best-estimate modeling capabilities. In this work, an SS-based tool, called NUSS is implemented which perturbs pointwise ACE-formatted nuclear data using multigroup nuclear data covariance. The use of pointwise ACE-formatted nuclear data in NUSS can accommodate flexible multigroup covariance structures and allows for nuclear data uncertainty propagation through the continuous/pointwise-energy transport code MCNPX. As a first step of the NUSS development and verification, uncertainty contributions from 239 Pu and 235 U nuclear data were assessed for Jezebel (Pu-fueled) and Godiva (U-fueled) fast-spectrum criticality benchmarks. NUSS results are compared to those by other uncertainty quantification methods such as TSUNAMI and MCNPX PERT CARD. Next, Light Water Reactor (LWR) pin cell models from the OECD/NEA UAM Phase-1 benchmark were analyzed. Results of cross section and k inf uncertainties in consideration of different nuclear data covariance libraries are presented. [ABSTRACT FROM AUTHOR]

Highlights: [•] Monte-Carlo MCNPX was used to analyze flux profile in a traveling wave reactor. [•] Results show steady propagation of flux (2cm/year) over life of the reactor. [•] High discharge burn-up of 394GWd/MTU was observed for the prototype compact model. [Copyright &y& Elsevier]

Highlights: [•] An atmospheric dispersion model for a hypothetical accident of Ghana Research Reactor-1 (GHARR-1) was developed. [•] Radiological safety analysis after the postulated accident was also carried out. [•] The MCNPX and HotSpot codes were used to achieve the objectives of our study. [•] All the values of effective dose obtained following the accident were far below the regulatory limits. [Copyright &y& Elsevier]

Highlights: [•] Review of several Monte Carlo effective delayed neutron fraction calculation methods. [•] These methods have been implemented with the Monte Carlo code MCNPX. [•] They have been benchmarked against against some critical and subcritical systems. [•] Several nuclear data libraries have been used. [ABSTRACT FROM AUTHOR]

Highlights: [•] Conceptual design study of fusion-fission hybrid reactor for waste transmutation. [•] MCNPX and MONTEBURNS are compared for transmutation performance of WT-Hyb. [•] Detailed neutronic performance of final optimized Hyb-WT design is analyzed. [•] A new tube-in-duct core design is implemented and compared with pin type design. [•] Study shows many aspects of hybrid reactor even though scope was limited to neutronic analysis. [ABSTRACT FROM AUTHOR]

Highlights: [•] 500kg Natural uranium target-blanket assembly irradiated with 1 and 4GeV deuterons. [•] Activation of Au and Bi samples used to study particle field of fast ADS. [•] Experimental results compared to Monte Carlo simulations using MCNPX 2.7 code. [•] Good agreement with experiment found for Bi reaction rates using INCL4-ABLA models. [•] CEM03 code showed better agreement than INCL4-ABLA model for 197Au(n,γ) reaction. [Copyright &y& Elsevier]

Highlights: [•] Among the kinetic parameters, the most important ones are βeff and Λ. [•] The kinetic parameters at the end of cycle for 20days operation at 1MW reactor power have been calculated using MCNPX code. [•] The system sub-criticality levels, the required accelerator current and power have been calculated for each burnup step. [•] Burnup effects on neutronic parameters and accelerator requirements are analysed for a life cycle of the reactor core. [Copyright &y& Elsevier]

Highlights: [•] We proposed a methodology to calculate importance weighted kinetic parameters. [•] Kinetic parameters of the TRIGA reactor have been calculated using MCNIC method. [•] The results of the method have been validated. [•] Parameter Λ is highly sensitive to the reactor geometry and reflector thickness. [•] MCNIC method can efficiently be used to calculate the kinetic parameters in ADSRs. [ABSTRACT FROM AUTHOR]

Highlights: [•] A 241Am–Be neutron source shielding system was optimized by simulating using MCNPX code. [•] The system was performed for boron compositions due to producing lower activated elements. [•] A compact shield was designed in terms of both moderating power and absorbing ability. [•] The arrangement was consistent with safety requirements and cost limitations. [•] Proper sites were determined in order to achieve maximum thermal and fast neutron flux. [ABSTRACT FROM AUTHOR]

Abstract: A relevant issue concerning the deployment of the very high temperature reactors is the accumulation of radioactive graphite and its final disposal. One way to reduce the amount of graphite in this type of reactors is to replace the reflector with other material. In this study the Gas Turbine Modular Helium Reactor was investigated, and most of the reflector was replaced with Ferritic–Martensitic steel. The Monte Carlo particle transport code MCNPX 2.6 was used. A fully detailed three dimensional core model was set up in order to evaluate the neutronic response of the modified core design. Results show that the proposed core performs well in terms of the neutron multiplication factor and reactivity coefficients. Considering a reactor life time of 60years, and taking into account 6years period for the replacement of the removable reflector, more than one thousand cubic meters of irradiated graphite are saved, which represents estimated economic saving of USD $237,541,000 and 7.73×1015 Bq in radioactivity. [Copyright &y& Elsevier]

Abstract: Recently, extensive research has been performed to produce neutron beams in hospitals using electron linear accelerator for neutron medical applications such as Boron Neutron Capture Therapy. In this article, we present the results of Monte Carlo calculations, using MCNPX code, for the possibility of setting up a 25MeV electron linac based photoneutron source for BNCT applications. Photon and photoneutron converters with different materials in various geometries were studied to achieve maximum neutron yield. In addition, the possibility of using an optimized neutron beam in BNCT was investigated by proposing the optimized beam shaping assembly. It was shown that our designed neutron beam, which passes through the optimized BSA, meets the IAEA criteria for in-air parameters. [Copyright &y& Elsevier]

Abstract: The prompt decay constant method and the area-ratio (Sjöstrand) method constitute the reference techniques for measuring the reactivity of a subcritical system using Pulsed Neutron Source experiments (PNS). However, different experiments have shown that in many cases it is necessary to apply corrections to the experimental results in order to take into account spectral and spatial effects. In these cases, the approach usually followed is to develop different specific correction procedures for each method. In this work we discuss the validity of prompt decay constant method and the area-ratio method in the Yalina-Booster subcritical assembly and propose a general correction procedure based on Monte Carlo simulations. [Copyright &y& Elsevier]

Abstract: The Supercritical Water Reactor (SCWR) is one of the most promising and innovative designs selected by the Generation IV International Forum. One of the concepts that is being studied is the High Performance Light Water Reactor (HPLWR) which is the European version of the SCWR. For this reactor, neutronics calculations have been performed with MCNP/MCNP5/MCNP4B/MCNPX4C and the results obtained are used as reference. However, MCNP calculations take too much computing time for design or fuel management analysis; therefore, it was decided to explore the factibility of using the HELIOS-2 code given that the computing time required for the simulations is notably reduced. The main objective of this work is to develop a model for the HPLWR fuel with HELIOS-2, and to subsequently compare the results against the MCNPX, in order to use the HELIOS-2 for performing further reactor physics studies given that although MCNPX is better HELIOS-2 is faster. Several variables, which are essential in the neutronics design of any reactor, were compared, that means: the infinite neutron multiplication factor (k ∞), the pin power distribution, and the power peaking factor for the fuel assembly, in hot and cold conditions, as well as the fuel reactivity coefficients. Calculations were made for a square fuel assembly with two rows of fuel pins, some of them containing gadolinia (Gd2O3) as burnable absorber material. The tests were simulated for different gadolinia enrichments as well as for no gadolinia. The calculations were done with the total reflection condition. The results show good similarities between the deterministic HELIOS-2 and the stochastic MCNPX calculations for the selected fuel assembly. Almost all values of the maximum power peaking factor obtained from MCNPX and HELIOS-2 are very similar to each other. Regarding the k ∞ values, the maximum relative difference (considering fuel temperature variations between 400 and 1200K) is 564pcm Δk/k for a coolant density of 0.74g/cm3 and 1200K. According to results, the HELIOS-2 code is good enough to be used in neutronics simulations of a HPLWR. In regards to reactivity coefficients, due to fuel temperature variations, we obtained similar results with both codes. We concluded that HELIOS-2 can be used for making HPLWR analysis faster than MCNPX. [Copyright &y& Elsevier]

Abstract: A 3D full-core heterogeneous model of the BFS-62-3A critical benchmark experiment was developed and validated using the Monte Carlo MCNPX-2.4.0 code. The BFS-2 critical facility at the Institute of Physics and Power Engineering (IPPE) was designed for simulation of fast reactor core neutronics, and for the validation of codes and nuclear data. The BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor core with (U,Pu) O2 fuel of 17% Pu content and stainless-steel reflectors. It was operated to measure the effective multiplication factor, spectral indices, radial fission rate distributions, control rod worths and sodium void effects. In the present study, special care was taken to run the MCNPX model to make Monte-Carlo confidence intervals comparable with uncertainties reported in the experiments; such as in material dimensions, number densities and isotopic compositions. In addition to the effective multiplication factor, sodium void effect, fission rate distributions and control rod worth were calculated. Simulations were carried out with four different modern nuclear data libraries; the primary aim being to estimate sensitivity of the results to the nuclear data. This task, besides being a library comparison, is also meant as a first step towards a code-to-code verification with deterministic methods. Results agree well with experimental values on most of the nuclear characteristics, even though a discrepancy up to more than 20% was found on the flux distribution in the stainless-steel reflector region, evaluated by measurements of the U235 and Pu239 fission rates; the sensitivity of this discrepancy to the stainless steel reflector density was estimated. [Copyright &y& Elsevier]

Abstract: The neutron fluence load on a reactor pressure vessel is an important physical quantity affecting material degradation and consequently reliable assessment of pressure vessel integrity and lifetime prolongation beyond designed conditions. This degradation is influenced by mixed neutron–photon source from the core periphery as well as by material parameters of the reactor pressure vessel. Computational procedures and experimental determination of neutron fluxes in the VVER-1000 mock-up internal structures as well as the results achieved are described in this paper. The calculation were performed with the MCNPX code with different nuclear data libraries. Nuclear data were processed using NJOY code. The neutron spectra measurements were performed with a two-parameter stilbene spectrometer. [Copyright &y& Elsevier]

Abstract: Two different types of perturbations of an SCWR-like fuel lattice have been investigated experimentally in the central test zone of the PROTEUS zero-power research reactor at the Paul Scherrer Institute in Switzerland. In each case, a campaign of high-resolution gamma-ray spectroscopy measurements was carried out on 34 fuel pins of the test lattice. In the first case, the test lattice was perturbed by inserting aluminum rods into the four central moderator regions, while in the second case, the perturbation was affected using steel absorber rods (instead of aluminum). The derived reaction rates are the capture rate in 238U (C 8) and the total fission rate (F tot), as also the reaction rate ratio C 8/F tot. Each of these has been mapped on the lattice and compared to calculated results from whole-reactor Monte Carlo simulations with MCNPX. Excellent agreement has been obtained, for both perturbed lattices, between the calculated and experimental distributions of C 8, F tot and C 8/F tot. Considering that control rods in an SCWR assembly are foreseen to be inserted into the central moderator regions, these results may be considered as generic validation of Monte Carlo simulations for the two different types of lattice perturbations which inserted control rods imply, viz. moderator displacement and strong neutron absorption. In a second step, calculated C 8, F tot and C 8/F tot distributions for the two perturbed lattices (as well as for the unperturbed lattice) have been compared, at assembly level, between MCNPX and the deterministic LWR lattice code CASMO-4E. In the case of the unperturbed lattice, as well as for the lattice with steel rods, the agreement between the codes is found to be within ∼1% for all pins and each reaction rate. However, for the lattice with aluminum rods, i.e. the case with mainly just moderator displacement involved, CASMO overestimates the reaction rates in the vicinity of the perturbations by up to 2–3%, when employing the standard input options. The reason for this discrepancy has been found to be the leakage treatment, which uses the fundamental-mode buckling applied in a homogenized sense across the lattice. In this way, global leakage gradients get averaged out over the entire assembly. The optional input card BZ2 for CASMO resolves this problem, and the codes then agree within 1% even for the aluminum case. [Copyright &y& Elsevier]

Abstract: Refined analysis, based on use of the Monte Carlo code MCNPX-2.4.0, is presented for the “H.B. Robinson-2 pressure vessel dosimetry benchmark”, which is a part of the Radiation Shielding and Dosimetry Experiments Database (SINBAD). First, the performance of the Monte Carlo methodology is reassessed relative to the reported deterministic results obtained with DORT. Thereby, the analysis is accompanied by a quantitative evaluation of the optimal energy cut-off value for each of the in- and ex-vessel dosimeters that were employed. Second, a more realistic definition of the neutron source is implemented than proposed in the benchmark. Thus, the current procedure for power-to-neutron-source-strength conversion, as also for explicitly considering the burnup-dependent fuel assembly-wise average fission neutron spectrum, is found to affect the calculated values significantly. In addition to the modelling refinements made, different approaches are tested for deriving the dosimeter activities, such that the neutron source time-evolution and the activity decay can be taken into account more accurately. Finally, in order to achieve a certain assessment of uncertainties, several sensitivity studies are carried out, e.g. with respect to the nuclear data used for the dosimeters, as also to the assumed physical location of the dosimeters. In spite of some apparent degradation in the prediction of experimental results when refining the Monte Carlo modelling, the final calculation/experiment (C/E) ratios for the measured dosimeter activities remain significantly improved (in general within 15% of unity), relative to the deterministic results. [Copyright &y& Elsevier]

Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained results are very similar. [Copyright &y& Elsevier]

Abstract: Monte Carlo simulation of a detector response function presents a very challenging problem. The detector response functions have been calculated for different neutron and gamma detectors: 3He gas filled proportional counter, NE213 organic scintillator, BrillanCe 350 or LaCl3(Tl), and an ionization chamber with mixed gas composition. MCNPX code was used for simulations. The simulations were done with different neutron and gamma energies. The effects of neutron scattering, wall effects, recoil continua and contribution from charged particles have been included. The detector response function for the NE213 organic scintillator was obtained with consideration of light output curves of different products of neutron reactions with materials of the scintillator. The simulated data has been compared with experiments. [ABSTRACT FROM AUTHOR]

Abstract: This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for ∼10% differences in the prediction of the minor actinide isotopes buildup. [Copyright &y& Elsevier]

Abstract: At the Paul Scherrer Institut, a methodology for PWR fast neutron fluence estimations, based on the Monte-Carlo particle transport code MCNPX with general-purpose neutron data libraries and using neutron source data from deterministic 3-D core-follow calculations, has been developed. The methodology has been validated on the basis of experimental data related to the fluence at the inner surface of a Swiss PWR Reactor Pressure Vessel. In this technical note, a first objective is to enlarge the validation basis of the methodology as well as to extend it for applications to RPV outer-surfaces. To that aim, a preliminary analysis with the MCNPX-2.4.0 code, along with the JEFF-3.1 continuous-energy neutron data library, of the “H.B. Robinson-2 Pressure Vessel Benchmark”, providing in-vessel and out-vessel experimental dosimetry data, is presented. In addition, considering that the available original solutions of the benchmark employed deterministic transport methods with associated libraries, a second objective of this technical note is to assess the progress achieved for this type of problems when applying modern Monte-Carlo based methods. The results show that for the 237Np and 238U fission dosimeters, which were of primary interest in the given study, a non-negligible improvement is seen in the MCNPX solution, indicating thus a rather good performance of the employed Monte-Carlo method and providing thereby, additional confidence for the overall PSI fluence methodology. For other high-energy dosimeters, the presented new results do not show yet any significant accuracy improvement versus previously reported results. This can however not be confirmed before additional studies, e.g., with focus on improvements of the involved modelling approximations and the statistical precision of the MCNPX calculations, be carried out. Similarly, investigations of neutron cross-section and dosimetry libraries effects remain to be addressed. These further studies are however not included here since at this stage, the principal aim was mainly to model and analyse this benchmark at most consistent manner with the previous solutions using a continuous-energy Monte-Carlo based method. [Copyright &y& Elsevier]

Abstract: The subcritical multiplication factor ks and the external neutron source efficiency are important parameters in the accelerator-driven system (ADS) performance assessment. The theoretical relation between ks and the effective multiplication factor keff in a subcritical system is discussed in different cases of subcritical system. On the basis of the theoretical background, the dependence of ks and on subcriticality, source position, and energy is numerically investigated using a simple thermal subcritical model. For the sake of experimental evaluation of ks and , the ADS experiments have been carried out in the subcritical systems combined with 14MeV pulsed neutrons of the Kyoto University Critical Assembly (KUCA). The ks and parameters are successfully measured by utilizing the reaction rate distribution obtained by the optical fiber detectors in the subcritical system, within a relative difference of less than 7% and 12% for ks and , respectively, between measured and calculated values for most studied cases. [Copyright &y& Elsevier]

• In-beam active neutron interrogation developed and deployed at the National Research Universal reactor. • MCNPX computational study completed for optimization of the neutron counting system. • The neutron counting system permits measurements of prompt and delayed neutrons using a mono-energetic and white thermal neutron beam. • The presence of a sub-milligram mass of 235U could be revealed in less than ten minutes. • The technique readily differentiated between 235U and 233U isotopes. • Detection time reduced to a few seconds in the examination of prompt neutrons. • Detection limits of fissile materials were calculated. A neutron counting (NC) technique for in-beam active neutron interrogation has been developed and deployed at the National Research Universal (NRU) reactor of the Canadian Nuclear Laboratories (CNL). A distinguishing feature of the method is that the sample to be interrogated remains stationary during the irradiation and counting periods for the detection of prompt and delayed neutrons. The NC system permits measurements of prompt and delayed neutrons using a mono-energetic or broad-spectrum (white) thermal neutron beam. An MCNPX computational study of an NC system employing nine 3He tubes showed that a delayed neutron detection efficiency of 22% could be achieved. The presence of a sub-milligram mass of 235U could be revealed in less than ten minutes in the in-beam delayed neutron experiment incorporating a white thermal neutron beam. The technique readily differentiated between 235U and 233U isotopes by analysis of delayed neutron count rates. The lower flux of the in-beam DN experiment does not permit the trace analysis that is possible with in-core irradiation, but does permit non-destructive analysis of large samples and could prove invaluable for the initial survey of materials of unknown content and origin. The study also demonstrated that in-beam prompt neutron analysis could be the ultimate solution when the neutron source is weak, the sample is shielded, the fissile mass is small, or interrogation time is limited. Detection time was reduced to a few seconds in the examination of prompt neutrons either in single or coincidence method. This paper presents the application of an in-beam NC system using a mono-energetic and white thermal neutron beam at the NRU reactor and assesses the performance of a newly constructed portable NC system with a large neutron detection array. [ABSTRACT FROM AUTHOR]

• A methodology for estimation of Tramp Uranium mass in BWR is considered. • Justification basis for the "Pu-based model" versus the "U-based model" is discussed. • Irradiation of Tramp Uranium on the outer clad surface is simulated. • Estimations on the effective fissile content and fission cross-section are provided. • Replacing the older U-based model by the Pu-based model is recommended. This paper presents calculation studies towards validation of a methodology for estimations of the mass of tramp uranium in a boiling water reactor (BWR) core from water chemistry measurements. The computational studies are carried out for a typical BWR fuel assembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclides and the fraction of 235U fissions to total fissions in different zones of a fuel rod, including tramp uranium on the clad surface, it is found that Pu gives the dominant contribution to fissions for tramp uranium after an irradiation on the outer clad surface of at least one cycle. Thus, the use of the so-called "Pu-based model" for the determination of the tramp uranium mass (this means in particular using the yields for 239Pu fission) appears justified in the cases considered. On that basis, replacing the older "U-based model" by the "Pu-based model" is recommended. [ABSTRACT FROM AUTHOR]