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Fission of light actinides has shown to exhibit features, which abruptly deviate from the monotonous trend established for heavier systems. More specifically, the description of their fission excitation functions and fragment angular anisotropies was found impossible to reconcile with the statistical- model framework employing a double-humped fission barrier profile, which was successfully applied to heavy nuclei. This was referred to as the "thorium anomaly" in the 1970's, and was interpreted in terms of the occurrence of a triple-humped barrier. Over the years, its impact on other fission quantities, in particular on the fission yields, has nevertheless been overlooked. In this work we show the drastic influence of the third barrier on the fission-fragment mass (equivalently, charge) and total kinetic energy at excitation energies close to the barrier. The analysis of these observables, and of their variation as a function of initial excitation energy, reveals the suppression of compact (and possibly very elongated) shapes at scission in light actinides. An extension of the semi- empirical GEF code is developed that relates this suppression to the influence of the third barrier. Accounting for this specific feature of light actinides is crucial for a proper description of the fission of the nuclides involved in 235 U-driven reactors, and, to a higher degree, those of the Th-U cycle. These features are not considered in present official nuclear-data tables. Systematic fragment-yield predictions are provided. • Low-energy fission of light actinides is influenced by the third fission barrier. • The resulting fission yields vary strongly as a function of excitation energy. • Current evaluations and nuclear-data tables do not consider this effect. • This finding is crucial for the design and the operation of fission reactors. [ABSTRACT FROM AUTHOR]

• A new time-dependent leakage model was developed for the operating conditions of VVER nuclear reactors for calculating the radioactivity of isotopes from nuclear fission and neutron-activated reactions both within the fuel and in the primary coolant. • The leakage functions for a complete reactor cycle can be determined by comparing the measured and the model calculated activities. Level of radioactive contamination in primary coolant of nuclear reactors must be kept as low as possible, therefore essential conditions is the leakage-free state of the fuel rods. To monitor the leakage defects in VVER reactors, the radioactivity of fission products must be measured in the primary loops. From radioactivity data it is possible to infer the fuel assemblies' hermetic or time-dependent leakage state. A new, time-dependent leakage model was developed for operating conditions of VVER nuclear reactors. This leakage model is able to calculate isotopes' radioactivity inside the fuel and in the primary coolant originating from nuclear fission and neutron-activated reactions. The core of the leakage model consists of several systems of differential equations describing dominant nuclear processes on time-dependent number of radioactive isotope nuclei within fuel assemblies and primary coolant. The model considers the effect of technological parameters on the activity of radioisotopes in the primary coolant. [ABSTRACT FROM AUTHOR]

• The possibility of a boiling water cooled travelling wave reactor is studied and confirmed. • Breeding is sufficient for reaching the criticality from depleted uranium, if the pressure is low. • Burn-up reaches up to 40% and the natural uranium utilisation is improved by 50 times. • 1-D numerical thermal hydraulics and neutronic burn-up coupled calculations are carried out. This study investigates the principal possibility of a boiling-water-cooled travelling wave reactor that can potentially improve the natural uranium utilisation in Boiling Water Reactors (BWRs), and more general in water-cooled reactors, appreciably. A strong variation of the water density along the core height makes favourable the nuclear fission in the lower core part and the fuel breeding in the upper one. A serial axial fuel shuffling scheme is considered for establishing a kind of travelling wave. The fresh natural uranium oxide fuel is loaded from the core top. The irradiated fuel is discharged from the core bottom. The water coolant at the saturation point at a near-atmospheric pressure comes from the bottom of the core, as in conventional BWRs. The asymptotic state of the breeding/burning wave is obtained, including its major characteristics, such as power density distribution, coolant mass flow rate, and the fuel shuffling speed. The employed simulation model for the water density variation in axial direction is based on a slip ratio approximation for two-phase flows. The burn-up calculations are performed with the ERANOS2.2 code, in which the models of axial fuel shuffling and coolant density variation have been implemented. The numerical results for 1D-models are encouraging and show that the breeding performance is sufficient to keep the core critical in the asymptotic state, and a maximum burn-up of about 37% can be reached, meaning that the natural uranium utilization in BWR can be improved by several ten times compared to conventional water-cooled reactors. Feasibility-related issues, such as the void reactivity effect and critical heat flux limit, has been checked and discussed. Further efforts are needed to establish a viable design based on the ideas proposed in the paper. [ABSTRACT FROM AUTHOR]

Highlights • Efficient, accurate, and stable coupling scheme for the aerosol dynamics in the containment. • Validation with the analytic solutions of the simplified general dynamic equations. • Analyses on international standard problem No. 44 to validate with the experimental data. • Negligible computation burden required for the coupling scheme as additional. Abstract Most fission products are released into the containment as aerosol particles during severe accidents. In the conventional codes, two sets of equations are used to analyze the aerosol behaviors in the containment. The first set of equations are multicomponent sectional equations, solutions of which are given with the mass concentrations in each discretized section. The multicomponent sectional equations require very large number of sections to remove the numerical diffusion when hygroscopic growth occurs with other aerosol dynamics simultaneously. For efficient calculations, Mason Equations are used to analyze the hygroscopic growth. The solutions of the equations, on the other hand, are given with the radius of an individual particle. Therefore, it is essential to couple the two equations in rigorous manner for efficient analyses on the aerosol behaviors when hygroscopic growth and other aerosol dynamics occur simultaneously. In this paper, we propose a new formulation for the coupling scheme of multicomponent sectional equations and Mason equations. In the formulation, the two equations are coupled via a transition rate matrix obtained by interpolation of the mass distribution within a section for each component in multicomponent sectional equations and the solutions of Mason equations. In order to validate the formulation, it is applied to the simplified GDE to compare with the analytic solutions. The formulation is also applied to international standard problem No. 44 for the validation with experimental data. Computation time for the formulation is also compared with those of the conventional method. With negligible computation burden as additional, the formulation would analyze the aerosol behaviors efficiently, accurately, and stably when hygroscopic growth occurs with other aerosol dynamics simultaneously. [ABSTRACT FROM AUTHOR]

Highlights • BR2 can perform simultaneous irradiation of DDE-NBSR & MITR as in original reactors. • MCNP BR2 burn-up model with DDE irradiation rigs is developed by BR2 management. • Requested targets for maximum heat flux & fission density are met for both DDEs. Abstract The BR2 reactor physics group has performed extensive neutronic feasibility studies of four Design Demonstration Element (DDE) tests foreseen for the US High Performance Research Reactor (USHPRR) conversion program. The purpose of these tests would be to irradiate the future LEU fuel elements for MURR, MITR, NBSR and HFIR in the BR2 reactor under conditions that are similar to those of the original reactor. The present paper focuses on the preliminary feasibility study for the simultaneous irradiation of the MITR-DDE and the NBSR-DDE in the BR2 reactor. In order to mimic the irradiation conditions of the original reactors for both DDEs inside the BR2 reactor, a number of irradiation scenarios and irradiation rig designs were optimized. The results obtained in this paper show that the BR2 reactor is capable to accurately recreate the typical irradiation conditions experienced by the future fuel elements of the NBSR and MITR with high fidelity. The power evolution and final burnup distributions are close to the specified values. We have proposed and have shown the possibility to irradiate both DDEs at the same time. The irradiations, if conducted as specified at fully representative powers, are long (three to five years), due to the requested relatively low power in combination with targeted elevated burnups. It is possible to reduce the irradiation time if a higher power can be accepted, but in this feasibility study, we attempted to simulate the required conditions as accurately as possible. [ABSTRACT FROM AUTHOR]

Highlights • DGM with a truncated basis was used to solve eigenvalue problems for the first time. • A POD-driven basis incorporate physics and leads to improved, truncated results. • Success of truncated expansions is largely independent of group structure. • Local fission densities within 0.01% of reference using just 6 DoF per coarse group. Abstract Presented is the evaluation of the discrete generalized multigroup (DGM) method using a truncated basis constructed from proper orthogonal decomposition (POD). The DGM method uses an orthogonal basis to collapse fine-group fluxes and cross sections to a coarse-group structure with higher order modes in each coarse-group. By solving the course-group equation, an approximation to the fine-group flux can be reconstructed from the course-group flux. This approximation is used to recollapse the cross sections, which leads to an improved approximation. Past development of the DGM method used the discrete Legendre polynomials (DLPs), which do not perform well under truncation, for approximating the spectral shape of the flux. In this work, basis sets constructed using POD were used for the DGM method for several fine-group structures. The POD basis incorporates spectral information from small, representative problems, which are designed to capture spectral information from the full problem. The most effective, practical POD basis approximated the fission density with an error of less than 0.01% relative to a reference problem using approximately 6 degrees of freedom per coarse-group for all fine-group structures. The reference chosen was a 1-D, repeating lattice of 10 pins of UO 2 fuel cells adjacent to 10 pins of 7.0% MOX fuel pins. The DGM method when subject to truncation performed far better using POD basis sets than when used with the DLP basis. [ABSTRACT FROM AUTHOR]

Abstract The Korea Atomic Energy Research Institute (KAERI) is planning the construction of the KIJANG Research Reactor (KJRR), which is currently undergoing licensing procedures. The fission 99Mo production process (FMPP) of the KJRR generates intermediate-level liquid waste (ILLW) and alumina residue (Al 2 O 3 cake), which is a solid waste. In the present study, an efficient method of reducing the amount of generated cement waste form by mixing the ILLW and Al 2 O 3 cake together, into low-level waste, was developed. The purpose of this study was to determine if the cement waste form produced by mixing cement with the waste from the FMPP would satisfy the acceptance criteria of a low- and intermediate-level waste disposal facility in Korea. The process operating range and the optimum operating conditions were found by investigating the mixability of the cement waste form, by considering the presence of free standing water after curing, and by performing a water immersion test depending on the mixing ratios of Portland cement (type I), ILLW, and Al 2 O 3 cake. The feasibility of treating the cement waste form produced under the optimum operating conditions was evaluated in relation to the acceptance criteria of a low- and intermediate-level waste disposal facility in Korea. The results showed that the cement waste form satisfied the acceptance criteria with regard to structural stability (compressive strength test, thermal cycling test, water immersion test, and irradiation test) and leachability (leaching test). [ABSTRACT FROM AUTHOR]

• Modeling the problem of FAL pattern optimization in PWRs using the TSR algorithm. • Increasing the multiplication factor in PWRs. • Flattening the power (reducing power peak factor) in PWRs. Pressurized water reactors (PWR) are a type of light water nuclear reactors that constitute most nuclear power plants in the world. In PWRs, the initial coolant is water that is pumped into the reactor core with high pressure where it is heated by the energy released from nuclear fission. Then, the high-pressure hot water flows into a steam generator where its thermal energy is transferred to the water with lower pressure, and steam is generated. The steam, then, spins turbines that rotate an electrical generator to generate power. It is crucial to increase the safety of PWRs in nuclear power plants. A measure to ensure the safety of these reactors is to optimize the fuel assemblies loading (FAL) pattern. This paper presents a method to optimize the FAL pattern of PWRs by the trees social relations (TSR) optimization algorithm. The TSR algorithm is a metaheuristic algorithm inspired by the hierarchical life of trees in the jungle. The proposed method models the problem of FAL pattern optimization by the TSR algorithm. This method is practically compared with other methods. Based on the results, it outperforms similar methods in terms of the multiplication factor, power flattening, and fitness of the final solution. [ABSTRACT FROM AUTHOR]

Highlights • An innovative method of source term assessment has been developed. • The model for core adiabatic heat-up stage in IAEA-TECDOC-1127 has been improved. • Four methods were used to comparatively study the source term released into the containment. Abstract In the event of severe accidents in nuclear power plants (NPPs), timely and accurate evaluation of the source term released from the reactor can provide technical support for a nuclear emergency response. Currently, the method for source term evaluation during a nuclear emergency in NPP is mainly based on the core damage assessment technique. The first step in the evaluation process is core damage assessment; and the second step is source term evaluation, based on core damage state and NUREG-1465 reference guide. In this paper, a nuclear emergency source term evaluation method (RVLST) based on measured reactor pressure vessel water level is presented. The single-step evaluation process combines pressure vessel water level monitoring system, core degradation model, and fission product release model. The feasibility, computational speed and accuracy of RVLST are compared with the methods in IAEA-TECDOC-955/NUREG-1465, IAEA-TECDOC-1127 and MELCOR using source term released into the containment in a large break loss of coolant accident (LBLOCA) in the AP1000 nuclear power plant, and the results obtained by the four methods were comparatively analyzed. The results show that the release fraction of noble gases calculated by RVLST is close to the one calculated by the IAEA-TECDOC-955/NUREG-1465 method and MELCOR. Moreover, compared with the results for Cesium (Cs) calculated with the IAEA-TECDOC-955/NUREG-1465 method, the calculations obtained from the RVLST are closer to those derived from MELCOR. Also, more conservative calculations were obtained from RVLST than those calculated with IAEA-TECDOC-955/NUREG-1465 and MELCOR for halogen Iodine (I) concentration. [ABSTRACT FROM AUTHOR]

Highlights • High-resolution time-of-flight neutron transmission measurements on enriched metallic samples of Mo-98 and Mo-100. • The Bayesian R-Matrix code SAMMY 8.0 was used to shape-fit the data and extract resonance parameters from the data. • The results represent a refinement of those given in the current evaluation due to improvements in experimental resolution. Abstract Molybdenum can exist in many nuclear reactor components, including fuel, cladding, or as a high yield fission product. As a result, accurate isotopic nuclear data for molybdenum are important for reactor simulation. To this end, high-resolution time-of-flight neutron transmission measurements on highly enriched isotopic metallic samples of Mo-98 and Mo-100 were performed and data were reduced to transmission in the resolved resonance region from 10 eV to 53 keV for Mo-98 and 10 eV to 26.5 keV for Mo-100. Measurements were taken with Li-6 glass transmission detectors at 31 m and 100 m flight paths. The Bayesian R-matrix code SAMMY 8.0 was used to shape-fit the data and to extract resonance parameters from the transmission spectra. The newly fitted resonance parameters were compared with those given in ENDF/B-VII.1. The results represent a refinement of those given in the current evaluation due to the improvement in the experimental resolution in these measurements. The comparison included analysis of level statistics. The resonance parameters for Mo-98 show many differences with the current evaluations. The results of the analysis indicated missing levels in Mo-98 starting at 10 keV, which implies an inability to resolve all resonances at higher energies. The resonance parameters for Mo-100 agree well with the current evaluation. Level statistics analysis indicates there are few missing levels up to 26.5 keV. [ABSTRACT FROM AUTHOR]

Highlights • Investigation of SBLOCA severe accident transient with ATHLET-CD code. • Mobile pump injection to reactor coolant system under various core conditions. • Assessment of mobile pump injection combined with primary side depressurization. • Quantification of releases of fission products and hydrogen. • Proposal of a preferred mobile pump injection strategy based on simulation results. Abstract This series of two papers focuses on the analysis of severe accident management measures for a generic German PWR of type Konvoi. A nuclear power plant model based on the severe accident code ATHLET-CD was developed in order to assess the code applicability for simulation of accident scenarios with core degradation. It was applied for investigation of two main groups of accident scenarios: station blackout and small-break loss-of-coolant accidents. Part I presented the analyses of a station blackout accident. Part II is focused on the analysis of the plant response in case of a hypothetical small-break loss-of-coolant accident (SBLOCA with 50 cm2 leak in cold leg of pressurizer loop). Due to failure of components of the emergency core cooling system and assumed unavailability of several preventive and mitigative accident management measures (AMM), the accident develops into a severe accident scenario with core melt and reactor pressure vessel failure. An alternative AMM, which is not implemented in German PWRs, is investigated in the current paper: late coolant injection to the primary circuit with a mobile equipment (bleed and feed with low pressure pump). The analysis is performed on the basis of ATHLET-CD simulations (code version Mod 3.0 Cycle A). Two injection points are investigated: injection to cold leg and injection to hot leg. As in SBLOCA scenarios the primary pressure might remain high that injection by low pressure systems is not effective, the influence of primary side depressurization (PSD) as additional AMM is investigated. The simulations show, that mobile pump injection significantly reduces the amount of the released fission products and hydrogen, if it is started within 75 min after the core exit temperature (CET) exceeded 400 °C. Without PSD, melting of the fuel can be prevented if the mobile pump is started immediately when CET > 650 °C, but release of fission products can only be prevented if early pump injection is combined with PSD. Cases with a delay of mobile pump injection in combination with a delay of PSD (e.g. due to temporary malfunction) are also investigated: until 50 min after CET > 400 °C a lower release of fission products and longer grace times to prevent melt relocation to the lower head are observed (compared to immediate PSD at CET > 400 °C). [ABSTRACT FROM AUTHOR]

Highlights • ATHLET-CD model for analysis of accidents with core degradation was elaborated. • Model application for severe accident analysis of a generic PWR of type Konvoi. • Analysis of accident management measures in case of station blackout. • Assessment of the effectiveness of the applied accident management measures. • Estimation of key timings and grace periods. Abstract This paper focuses on analysis of severe accident management measures for a generic German PWR of type Konvoi. A nuclear power plant model based on the severe accident code ATHLET-CD was developed in order to assess the code applicability for simulation of accident scenarios with core degradation. It was applied for investigation of two main groups of accident scenarios: station blackout and small-break loss-of-coolant accident. Part I of series of two papers analyses the plant response in case of hypothetical station blackout severe accident. Assessment of accident management measures in the preventive and in the mitigative domain is performed, where a focus is given on the combination of primary pressure reduction and injection by portable equipment directly into the reactor circuit. Key timings for operator actions are deduced. Both positive and negative effects of the investigated accident management measures are discussed. The results from the station blackout simulations showed that the time until core degradation can be delayed by application of primary side depressurization and usage of mobile pump as accident management measures. Depending on the time of injection significant reduction of the hydrogen and fission products releases can be obtained. In case that early water injection is possible, severe core damage might be prevented. [ABSTRACT FROM AUTHOR]

Highlights • Developed a detailed Zeus/Comet critical assembly model for IER-163. • Evaluated discrepancies between experimental and modeled results. • Nuclear data and experimental uncertainties were propagated through several methods. • The Monte-Carlo methods provide a thorough estimate of uncertainty for the indices. Abstract This paper describes several approaches to propagating nuclear data uncertainty in integral capture and fission reaction rates, focusing on the spectral hardness indices of 193Ir(n,n′)193mIr/191Ir(n,g)192Ir and 238U(n,f)/235U(n,f) in fast critical assemblies. Uncertainties in the nuclear data were propagated through the following methods: a sensitivity method (utilizing first order perturbation and covariance matrices), a BFMC method (random cross sections generated through covariance matrices), and Total Monte Carlo method (random cross sections generated from varying nuclear parameters). The uncertainty estimates from the three approaches are evaluated against the associated experimental uncertainties of a ZEUS/COMET critical assembly irradiation. The uncertainty estimates of the three approaches agree fairly well, and contribute a significant portion to the overall uncertainty of the spectral hardness indices. [ABSTRACT FROM AUTHOR]

The EURATOM FP7 project FREYA was launched in 2011 to support the design and licensing of lead-cooled reactors, by choosing as reference the MYRRHA and ALFRED fast spectrum systems. During the five-year project, a number of critical experiments were conducted in the VENUS-F zero-power reactor with the main aim to validate the neutronic codes adopted for the core design. Five critical core configurations were assembled in VENUS-F: the start-up core, the mock-up dedicated to ALFRED and three layouts simulating the MYRRHA reactor conditions. The ERANOS code system, coupled with the JEFF-3.1 nuclear data library, was one of the tools adopted for the design and characterisation of the experiments. In this paper, the results obtained with the ERANOS deterministic code are compared with the experimental measurements of integral and local parameters, such as: the core reactivity and kinetic parameters, the lead void reactivity worth, the axial and radial distributions of fission rates for the nuclides of major interest, the spectral indexes of important actinides ( e.g. , U 238 , Pu 239 , Np 237 , Am 241 ) respect to U 235 . The comparison between measurements and simulations shows a general good agreement, with some discrepancies for few local parameters that are here discussed. [ABSTRACT FROM AUTHOR]

When criticality calculation is performed with Reactor Monte Carlo (RMC) code for reactor analysis, the fission source iterative method is selected. There are two strategies to sample the fission neutrons in each fissile region: the conventional Proportional Stratified Sampling Method that focuses on the real physical process, and the author-developed Optimal Stratified Sampling Method that focuses on the optimal mathematics treatment. The Optimal Stratified Sampling Method have the effect of global variance reduction and could be easily implemented with the use of optimal source bias coefficients. However, the calculation of optimal source bias coefficients needs the values of variances in advance, which is almost impossible. To meet the requirement, a Batch Approximate Method is then put forward. The Batch Approximate Method based on the Batch algorithm and Shannon Entropy Diagnosis, eliminates the correlation among cycles during the criticality calculation. Combining the Optimal Stratified Sampling Method and the Batch Approximate Method, the Optimal Source Bias Method is established and developed in RMC code. The Optimal Source Bias Method was tested with a standard Pressurized Water Reactor assembly. The effect of global variance reduction was found, as well as the flattening of variance distribution. So, the new method is valuable for the high fidelity computation of the RMC code. [ABSTRACT FROM AUTHOR]

This work investigates the depletion capability implemented in lattice physics code STREAM for the prediction of pressurized water reactor (PWR) uranium dioxide (UO 2 ) spent nuclear fuel (SNF) isotopic inventory. The validation of this capability is performed by comparison of STREAM calculation results to measured SNF assay data obtained from PWRs Takahama-3, Calvert Cliffs and GKN II. The depletion analysis is conducted with the ENDF/B-VII.0 library and uses a pin cell model of the fuel rods from which the fuel samples were taken. The Chebyshev Rational Approximation Method (CRAM) is used to solve the depletion equation with about 1300–1600 isotopes in the depletion chain. 16 actinides and 23 fission products are analyzed in 14 spent UO 2 fuel samples. The actinides are isotopes of uranium, neptunium, plutonium, americium and curium. The fission products nuclides include isotopes of cesium, neodymium, europium, samarium as well as 106 Ru, 144 Ce, 155 Gd, 99 Tc, 90 Sr, 109 Ag, and 103 Rh. The sensitivity of some of the nuclides to the details of the power history and the adjustment of the fuel sample burnup is discussed. The impact of using ENDF/B-VII.0 library instead of ENDF/B-VI.8 is also discussed. Most of the nuclides analyzed are well predicted within ±7% of the experiment for actinides and fission products. STREAM depletion results are also compared to the codes SWAT, HELIOS and SCALE results based on publicly available information in literature, to check the performance of STREAM relative to other codes for the prediction of SNF isotopic inventory. The comparison to other code systems shows that the implementation in STREAM is of comparable accuracy. Overall, this paper demonstrates that the depletion capability in STREAM can be reliably applied to predict the isotopic inventory of PWR UO 2 SNF for burnup ranging from 14 to 54 GWd/t and initial enrichment ranging from 3.0 to 4.1 wt% 235 U. [ABSTRACT FROM AUTHOR]

Cyclic voltammetry (CV) analysis was performed to determine the behavior of Dysprosium (Dy) and Lanthanum (La) in molten FLiNaK salt (LiF-NaF-KF: 46.5–11.5–42 mol%) and to understand the mechanism driving the electrochemical reactions. The goals of these experiments are to understand the behavior of contaminants and nuclear fission products related to the Fluoride Salt-cooled High-Temperature Reactor (FHR) concept, and to generate a reliable source of electrochemical properties in fluoride salts. The tests were performed at 600–650–700 °C for Dy and 650–700–750 °C for La in a glassy carbon crucible contained within an inert argon atmosphere. A platinum wire quasi-reference electrode exhibited stable potential throughout experiments and allowed for thermodynamic reference between test species and known redox couple (F 2 /F − ). The electrochemical reduction of Dy and La (assumed three-electron, single-step transfer) maintained quasi-reversibility at the solid tungsten electrode for scan rates of 160 mV/s and below during CV. For scan rates of 200 mV/s and above, significant deviation from reversibility was evidenced by increasing separation between anodic and cathodic peaks. For Dy 3+ , the diffusion coefficient was calculated as D [cm 2 /s] = 0.0538∗exp(−6193/T(K)) and the activation energy E a  = 51.5 kJ/mol. The diffusion coefficient of La 3+ was calculated as D [cm 2 /s] = 14.252∗exp(−15285/T(K)) and the activation energy E a  = 127 kJ/mol. The apparent potential of the La 3+ /La 0 redox reaction was determined as E 0∗ [V vs. K + /K] = −0.368 + 0.001119∗T[K] and for the Dy 3+ /Dy 0 reaction as E 0∗ [V vs. K + /K] = 0.0626 + 0.00088∗T[K]. [ABSTRACT FROM AUTHOR]

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239 Pu and 240 Pu. The sensitive benchmark parameters that in this work represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport. [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]

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m 2 ) whereas the first type provides the lowest heat fluxes (around 500 MW/m 2 ) but is not realistic due to the non-miscibility of steel with UO 2 . Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m 2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters). [ABSTRACT FROM AUTHOR]

Aiming at developing the activation method and provide reference for the validation of thorium fission parameters, a dedicated integral experiment was carried out in a thorium oxide cylinder bombard by D-T fusion neutrons. Thorium fission rate in the axial direction of the cylinder was determined by detecting the 1260.409 keV gamma emitted from fission product 135 I, with experimental uncertainties of 6.2%. MCNP calculations employing ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0, CENDL-3.1 library data were in good agreement with measured thorium fission rates within experimental uncertainties. The agreement of measured reaction rates of monitors with the calculations on the whole demonstrated the reliability of this experiment in some degree, and could provide reference for the analyses of thorium fission rates. Neutron spectra and the contribution of neutrons in each sub-energy range to thorium fission rate were also analyzed. The activation method to determine thorium fission rate was developed and the data obtained in this work could provide reference for the validation of thorium fission parameters. [ABSTRACT FROM AUTHOR]

A new feature for the core calculation of prismatic-block type of Very High Temperature Reactor (VHTR) was added to our in-house code NuCoT. The nodal method used was Hybrid Nodal Green’s Function Method (HNGFM). The conformal mapping method was adopted for handling the hexagonal geometry. The accuracy of the new capability was verified through several benchmark problems. Results of comparable accuracy with other developments were obtained. The method was further verified by performing a 2D simplified HTTR problem and compared with MCNP result. The statistical errors of k eff was <10 pcm and of fission power was <0.1%. The results of Hex-capability of NuCoT showed that using supercell method for cross section generation for homogenization of fuel block gave more accurate results than using single fuel block with periodic boundary condition. For the control rod block that contains three large void channels, using anisotropic diffusion coefficient resulted in more accurate k eff . In the all-rods-out condition and the partially-controlled condition, the difference in k eff was <140 pcm and in maximum power was <0.8%. In the all-rods-in condition, the error in k eff and maximum power error were 413 pcm and 1.1%, respectively. [ABSTRACT FROM AUTHOR]

The subcritical multiplication factor k s with the existence of external source neutrons is important for the evaluation of the neutron multiplication M , especially in the accelerator-driven system (ADS) performance assessment. The effective multiplication factor k eff calculated from the traditional source iteration method in Monte Carlo codes can’t fully describe the subcritical system, acquiring the spurious neutron flux distribution. A direct Monte Carlo method called modified source iteration method by external source is introduced to calculate the subcritical multiplication factor k s directly, acquiring the real neutron flux at the same time. Compared with Monte Carlo fixed source calculation, the modified source iteration method has a great advantage in calculating the subcritical multiplication factor k s and its statistical error directly, especially for systems with complex geometry and a variety of fissile materials where much post-processing work on the tally results from fixed source calculation is needed. Moreover, much calculation time can be saved through this method when obtaining the subcritical multiplication factor k s with similar precision. This method has already been implemented in Reactor Monte Carlo code RMC including parallel calculation capability and can be effectively applied to the analysis of accelerator driven subcritical system (ADS). [ABSTRACT FROM AUTHOR]

The analysis of 239 Pu was performed using the SAMMY Code for Neutron Transmission, Capture, and Fission cross-sections in the resolved resonance region up to 2.5 keV. Due to the difficulties of processing large covariance matrices, the evaluations were performed in the following distinct uncorrelated energy regions: 0.0–1.0 keV, 1.0 keV–2.0 keV, and 2.0 keV–2.5 keV. By using version 8.0.0 of the SAMMY code, the treatment of large resonance parameter covariance matrices was improved significantly. The aim of this work is to analyze the experimental data using a self-developed experimental nuclear data fitting code FITWR and analyze it with the SAMMY code, then compare the results with Bayes method. Finally, the output results from these methods is processed by the SAMMY code in order to obtain covariance matrix resonance parameters in the energy range from 0 to 2.5 keV. All the computational experiments done for this paper was carried out at IMAN1 Center – Jordan’s National Supercomputing Center (IMAN1). [ABSTRACT FROM AUTHOR]

The Fluoride Salt Cooled High Temperature Reactor (FHR) is a liquid salt-cooled reactor utilizing TRISO-based fuel. Its low operating pressure and large thermal margin offer attractive safety. However, its complex core design, in particular the double heterogeneity introduced by TRISO particles, as well as the presence of several moderating materials, require careful verification and validation (V&V) of the core physics modeling and simulation (M&S) methods. Due to their ability to accurately simulate complex geometries and utilize continuous-energy nuclear data, Monte Carlo methods are used in this study to evaluate core physics phenomena and provide reference solutions for verification of lower fidelity methods. This work is focused on the FHR design with hexagonal fuel elements consisting of fuel plates (“planks”), i.e., the Advanced High-Temperature Reactor (AHTR) developed by ORNL. The pebble-bed FHR design, PB-FHR, is not directly considered in this work, although the results of this study are relevant for PB-FHR as well. The SERPENT Monte Carlo code is selected and utilized to perform core physics analyses of an FHR fuel assembly. The issues investigated include detailed fission distribution, quantification of certain approximations, study of neutron spectra in different material regions, and impact of graphite density, temperature, and scattering matrix data. This study contributes to V&V of modeling and simulation methods for FHR. [ABSTRACT FROM AUTHOR]

Ex-core neutron flux monitoring system (ENFMS) is designed to control nuclear reactor core power, its change rate and reactivity. Problem of core power evaluation by readings of the ENFMS is considered in this paper. Readings of ex-core ionisation chambers is main information source of the ENFMS on core state. But the ex-core ionisation chambers readings aren’t the only information source on core power level. Control rods positions, fuel burnup and several other physical parameters characterizing the state of the core have an impact on the ex-core ionisation chambers readings. The ENFMS has to take into account these parameters for evaluation of the core power with high precision. Power reading correction method (PRCM) for evaluation of the core power and other power parameters by the ENFMS using additional in-core information is discussed in this paper. Besides the ENFMS detectors readings this method takes into consideration control rods positions, coolant properties in downcomer region, xenon distribution, fuel burnup and decay heat power. Model tests demonstrated that the PRCM allows evaluating the power level and other power parameters of interest with high precision. Power evaluation error is reduced by factor 2–3 as compared to traditional method. [ABSTRACT FROM AUTHOR]

The possibility of computing adjoint-weighted scores by Monte Carlo methods is a subject of active research. In this respect, a major breakthrough has been achieved thanks to the rediscovery of the so-called Iterated Fission Probability (IFP) method, which basically maps the calculation of the adjoint neutron flux into that of the neutron importance function. Based on IFP, we have recently developed the calculation of effective kinetics parameters and sensitivity coefficients to integral reactor responses in the Monte Carlo production code T ripoli -4®. In view of the next release of the code, we have added a new routine allowing for the calculation of the adjoint angular flux (and more generally adjoint-weighted sources) in eigenvalue problems, which can be useful for code-code comparisons with respect to deterministic solvers. In this work we analyse the behaviour of the adjoint angular flux as a function of space, energy and angle for a few benchmark configurations, ranging from mono-kinetic transport in one-dimensional systems to continuous-energy transport in fuel assemblies. The Monte Carlo adjoint flux profiles are contrasted to reference curves, where available, and to simulation results obtained from ERANOS and APOLLO2 deterministic codes. [ABSTRACT FROM AUTHOR]

This paper presents an application of the hybrid transport point kinetic (HTPK) technique to the reactivity determination in subcritical reactor configurations. The mathematical model of the HTPK, initially proposed by Picca et al. (2011) to simulate the time-dependent neutron transport, is here extended to incorporate delayed emissions. The classical area method (Sjӧstrand, 1956), developed to invert the point kinetic (PK) model, is then adapted to accommodate the peculiarities of the HTPK approach, to allow its analytical inversion. This novel inverse neutron kinetic methodology is tested on a three-region reactor configuration, showing the interesting performance of the approach based on the HTPK model as compared to the standard area method and highlighting its potential to overcome some of the limitations of the PK-based inversion. [ABSTRACT FROM AUTHOR]

New reactor concept studies, such as those of the MYRRHA Accelerator Driven System (ADS), with their extended nuclear data needs relaunched the interest in sensitivity and uncertainty (S/U) studies. S/U analysis play an essential role in assuring the accuracy and reliability of nuclear reactor computations and provide an insight in the physical phenomena involved in the neutron transport. This paper presents the study of the MYRRHA critical core, which is designed to operate at 100 MW thermal power. The previously presented analysis of the MYRRHA effective multiplication factor ( k eff ) was extended to study the sensitivity and uncertainty in kinetic parameters, such as the effective delayed neutron fraction ( β eff ). The sensitivity and uncertainty computations were performed by means of the SUSD3D code based on forward and adjoint-flux first-order perturbation theory. The sensitivity coefficients of β eff with respect to the basic nuclear data were calculated by the k-ratio derivation method proposed in 2011. The main components of the uncertainties in β eff were estimated using the JENDL-4.0u covariance data. The energy dependent sensitivity profiles of k eff and β eff with respect to nuclear data are compared in view of exploiting the differences among them for a physically more complete, comprehensive and consistent nuclear data validation and improvement procedure. The accuracy of the k eff and β eff measurements and calculations has to be carefully evaluated in order to conclude on the nuclear data status and possible improvements. [ABSTRACT FROM AUTHOR]

This paper presents the theory and application of a code called CALLISTO which is used for performing NPP start-up and power ascension calculations. The CALLISTO code is designed to calculate various values relating to the neutron population of a nuclear system which contains a low number of neutrons. These variables include the moments of the PDF of the neutron population, the maturity time and the source multiplier. The code itself is based upon the mathematics presented in another paper and utilises representations of the neutron population which are independent of both space and angle but allows for the specification of an arbitrary number of energy groups. Five examples of the use of the code are presented. Comparison is performed against results found in the literature and the degree of agreement is discussed. In general the agreement is found to be good and, where it is not, plausible explanations for discrepancies are presented. The final two cases presented examine the effect of the number of neutron groups included and finds that, for the systems simulated, there is no significant difference in the key results of the code. [ABSTRACT FROM AUTHOR]

This paper focuses on the source term estimation in AP1000 Nuclear Power Plant (NPP) under severe accident condition. Methods that utilize MELCOR code can evaluate the source term, but the evaluation process is time-consuming. Also, the reference values in NUREG-1465 can be directly used as the severe accident source term; however, it was initially proposed for second–generation reactors and it has difficulty in simulating changes in release fraction along with accident process. In this paper, the Improved Simplified Source Term (ISST) code was developed based on the IAEA’s Simplified Source Term model (SST) presented in IAEA-TECDOC-1127. In addition, the models for core uncovery stage and adiabatic heating-up stage were optimized in this work; taking into account the Accumulator (ACC) and the Core Makeup Tank (CMT) of the emergency core cooling system. Moreover, four fission products release evaluation models were integrated into ISST, making it applicable for the source term evaluation of AP1000 reactor under severe accident initiated by LBLOCA condition. To verify the developed ISST code, the IAEA’s SST, MELCOR and the reference source term in NUREG-1465 were differently used to evaluate the source term released into the containment in a LBLOCA of the AP1000 nuclear power plant. The results obtained from the models in the ISST code were analyzed and compared with each other, and with those obtained from NUREG-1465, IAEA’s SST and MELCOR. Compared with results calculated with the IAEA’s SST, the results calculated with ISST are closer to those calculated using MELCOR. ISST is computationally faster than MELCOR, hence it gives the source term calculation results quicker. Compared with results calculated with MELCOR, the results calculated with ISST are more conservative for Cs and I while the result derived from NUREG-1465 for noble gases is more conservative. [ABSTRACT FROM AUTHOR]

Appraising the credibility and competency of a computer simulation can be done by verification and validation procedure. This study has been performed to validate SuperMC for precise simulation of metal-cooled fast reactors using a high profile heterogeneous critical core, the BFS-62-3A. This critical assembly – being attractive because of its multi-zoned structure with MOX fuel, high axial and radial heterogeneity, diversity of materials and their compositions, and eligibility/feasibility to benchmark the experiments and calculations – is a full scale mock-up of BN-600 hybrid core with different fuel regions and stainless steel reflector. Modeling of complex core design, calculation of desired parameters, and visualization of the complete model were all performed using SuperMC. The deviations of less than 0.4% and 0.8% appeared in the calculated values of multiplication factor and spectral indices respectively while average discrepancies for radial fission rate distributions in the fuel region and control rod worths were found to be 2.7% and less than 5% respectively that showed a good agreement of the simulation results with benchmark’s measured data. Present work thus validated SuperMC’s capability to precisely model a complex heterogeneous reactor core and to accurately calculate various neutron physics parameters that ultimately rendered a high level of confidence on its use in fast-spectrum and MOX-fueled nuclear fission reactors. [ABSTRACT FROM AUTHOR]

The Fukushima accident on March 11, 2011 has shown the relevance of examinations of severe accident inside a spent fuel pool (SFP) during beyond-design-basis external events, and the necessity for provisions to cope effectively with such events through a relevant severe accident management (SAM) strategy. Although the low decay heat of fuel assemblies and the considerable water inventory in an SFP can slow the progress of an accident compared to an accident in the reactor core, the numerous number of fuel assemblies stored inside it and the fact that the SFP building is not leak-tight present the potential for the formation of a direct path for fission products to rise from the SFP into the environment (i.e., a much greater severe accident risk). The purpose of this paper is to assess the effectiveness and success conditions of an emergency makeup water injection strategy, which is being as a representative SFP SAM measure after the Fukushima accident. Two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) were considered in the analysis. For the foregoing SAM strategies, the analysis results and relevant insights are summarized in relation to two major aspects: (a) the key events of the progression of an accident (such as the exposure, heat-up and degradation of the fuel assemblies; the generation of combustible gases such as Hydrogen; and the over-pressurization of the SFP building) and (b) the release of radiological fission products (such as Cesium and Iodine) into the environment. A simulation tool for severe accidents, MELCOR1.8.6, was used in the present analysis. [ABSTRACT FROM AUTHOR]

A Sodium-cooled Fast Reactor (SFR) as a Generation IV (Gen-IV) advanced type of nuclear fission energy requires a fuel transfer machine (or a refueling system). It is employed to exchange spent fuel-assemblies during plant operation and other failed core assemblies (CAs). The refueling system provides a means of transporting, storing and handling reactor CAs. For the refueling system of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), developed by the Korea Atomic Energy Research Institute (KAERI), the PGSFR refueling mechanism was modeled in this paper for the refueling planning, and its kinematics analysis (mathematical steps for forward and inverse kinematics) revealed interesting results by a simple and efficient 2D four-bar linkage system as a kinematics model. This paper argues that restricting coupler link movements is far better not only for more suitable solution forms but also for dynamic safety guarding against collisions with neighboring structures than setting an orientation of the gripper. Computer simulation results in all the angular position information based on the proposed approach report that the total rotation angles of the gripper link for all the core-to-core locations is less than 90° by restricting the coupler link. It was concluded that refueling with the middle component motion restricted is a safe operation strategy with other computational benefits. By taking advantages of the proposed approach, the refueling system can be ready for more precise and safe operations confirmed by kinematics solutions for all of the target core-assembly locations. [ABSTRACT FROM AUTHOR]

Misloading of a spent fuel assembly could occur during the loading process of transportation casks due to inaccurate burnup records or mistaken assembly identification. Therefore, criticality safety analysis is important for such conditions to ensure that the cask remains subcritical with a sufficient margin. Investigations of assembly misload in PWR spent fuel cask has been conducted in previous studies, a similar analysis is still needed for BWR casks because of the differences between them, such as operational conditions, assembly size, gadolinium rods presence, enrichment variation, and others. Consequently, a quantitative computational criticality safety analysis is conducted in this work for a variety of misloading conditions in a BWR burnup credit shipping cask (GBC-68). The uncertainty in k eff estimation due to both nuclear data and statistical sampling has been quantified to ensure that k eff uncertainty margin is within criticality bounds. The analysis has been carried out using different modules within the SCALE code system, namely, TRITON, KENO-V.a, and TSUNAMI-3D. The results demonstrate that BWR misloading has different trend than PWR because of the presence of gadolinium rods. Two sets of nuclides have been considered in this analysis to show the credit of the fission products: major actinides and major actinides plus fission products. The peak reactivity burnup is calculated to be 24 GWD/MTU. Assembly misload results illustrate that misloaded assemblies with burnup less than the peak reactivity burnup insert more reactivity than the fresh misloaded assemblies, which is the opposite for PWR where misloaded fresh assemblies insert more reactivity than the burned assemblies. The uncertainty analysis demonstrates that reactivity uncertainty is within 5% for most of the cases considered in this analysis. The estimation of probability of occurrence of misloading accidents is out of scope of this work and it is left as a future work. [ABSTRACT FROM AUTHOR]

Within the bilateral project between the CEA Cadarache and the Jožef Stefan Institute (JSI) a wide variety of measurements using multiple fission chambers simultaneously inside the reactor core were performed. The fission rate axial profiles were measured at different positions in the reactor core and at different control rod configurations. A relative comparison of the calculated fission rates using the MCNP code and the measured fission rates was performed. In general the agreement between the measurements and calculations is good, with the deviations within the uncertainties. For better observation and understanding of the neutron flux redistribution due to the control rod movement, the neutron flux and fission rate had been calculated through the entire reactor core for different control rod configurations. The detector position with minimum signal variations due to the regulating and compensating control rod movement during normal operation was determined. The minimum variation is optimal in case we want to reliably determine the reactor power without influence of the regulating and compensating control rod positions. [ABSTRACT FROM AUTHOR]

Electrorefiner (ER) is the heart of pyroprocessing technology which contains different fission, rare-earth, and transuranic chloride compositions during the operation. This is still a developing technology that needs to be advanced for the commercial reprocessing design of used nuclear fuel (UNF) in terms of intelligent materials detection and accountability towards safeguards. A novel signal detection, artificial neural network (ANN), has been proposed in this study to apply on massive ER systemic parameters to simulate cyclic voltammetry (CV) graphs for the unseen situation. ANN could be trained to mimic the system by driving the data sets interrelation between variables to provide current and potential simulated data sets with a high accuracy of prediction. For this purpose, over 230,000 experimental data points reported in literature have been explored—0.5–5 wt% of zirconium chloride (ZrCl 4 ) in LiCl-KCl molten salt with different scan rates at 773 K. This study has illustrated a new framework of ANN implementation to eliminate trial and error approach by comparing the average error of one to three hidden layers with different number of neurons. In addition, this framework results in finding a preferable balance between underfitting and overfitting in deep learning. Furthermore, simulated CV graphs were compared with the experimental data and illustrated a reasonable prediction. The results reveal two structures with three hidden layers providing a good prediction with a low average error. The outcomes indicate that ANN has a strong potential in applying toward safeguards for pyroprocessing technology. [ABSTRACT FROM AUTHOR]

A novel nuclear reactor concept is presented in this paper which can sustain the controlled fission reaction and has an innovative mechanism of reactivity control. The concept of this reactor is based on the stability of colloidal suspensions. These suspensions consist of colloidal fuel particles suspended in a moderator with a controllable separation distance. The packing factor of the colloidal suspension can be controlled by slightly modifying the ionic concentration of the suspension medium, effectively controlling the criticality of the reactor. Based on local average separation distance between particles, thermalization of neutrons can be varied and can be used to manipulate neutron multiplication factor k ∞ . Numerical calculations are performed with the help of MCNP software to obtain optimal configuration of critical assemblies. For the numerical case studies presented in this paper, uranium dioxide ( UO 2 ) fuel particles with size ⩾10 μ m in different liquid solvents are considered. The fuel particles are simulated as hard spheres packed in Bravais lattice structures within the solvent medium. Simulations involve neutron interactions of a uniform colloidal suspension of spherical fuel elements of diameters 5 cm–0.001 cm. The homogeneous fuel equivalence was found to occur for particles sized below 0.010 cm diameter. Criticality curves are generated for fuel particle configurations with varying particle density, confirming fuel configurations can be controlled by only modifying particle packing factor. Additional criticality curves are generated for various cylindrical geometries of fuel suspension vessels with modular form factors. Triethylene glycol is substituted as a moderator and suspension fluid with higher boiling point as compared to water. The type of moderator or solvent does not have any significant impact on criticality performance. [ABSTRACT FROM AUTHOR]

SuperMC is a general, intelligent, accurate and precise simulation software system for the nuclear design and safety evaluation of nuclear systems. In this study, the latest version of SuperMC has been validated by Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS), which is a full-core commercial Westinghouse PWR core, including two cycles of measured operational data. The benchmark is widely used in the validation of Monte Carlo codes, because it is an extremely detailed, relevant and large-scale 3D PWR model based on a real reactor. Firstly, the detailed BEAVRS model is built with the CAD-based hierarchical modeling function of SuperMC. Secondly, several calculations using SuperMC at hot zero power condition of cycle 1 are performed, and the comparison with the reference results and measured data show great agreement, including neutron effective multiplication factor, control rod bank worth, temperature coefficients, U-235 fission rates and relative power distribution. This study demonstrates the capability of SuperMC on 3D full-core high fidelity reactor simulation. [ABSTRACT FROM AUTHOR]

In this work we present a methodology of solution of the multigroup multi-layer stationary neutron diffusion equation in two-dimensional cartesian geometry. This eigenvalue problem describes the criticality of nuclear reactor, that is, it establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve this problem, we use the power method to obtain the dominant eigenvalue ( K eff ) and its corresponding eigenfunction. Each iteration of the power method requires the solution of a non–homogeneous diffusion problem, that usually is solved numerically, however in this work the neutron diffusion equation is solved in analytical form in each iteration. To solve this system of second order partial differential equations, we propose to use the Finite Fourier Transform in one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. After it is solved, we use the inverse Fourier Transform to reconstruct the expression of the neutron flux in the original variables. However, at each iteration of the power method it is necessary to update the source term with the neutron flux and the K eff of the previous iteration. Thus in all iterations new terms are added which becomes the process very laborious. To overcome this problem, the authors propose a methodology that approximates the neutron flux in standard form by polynomial interpolation. In order to reduce computational time we propose to subdivide the real regions of the problem into small fictitious regions. In this way, the interpolating polynomials of each region can be of low order, reducing the dimensions of the matrices involved and, consequently, computational time. The methodology is implemented to solve a heterogeneous problem and the numerical results are compared with the finite volumes method. [ABSTRACT FROM AUTHOR]

Reactor noise, caused both by the probabilistic nature of the fission chains and external reactivity noises, is one of the basic topics in nuclear science and engineering, both in theory and practice. Classical approaches to modeling this noise and neutron count distribution in the detection system rely on the stochastic transport equation for the probability generating function and on transfer function response to random perturbations. In recent years, a third modeling approach has been proposed, relying on Ito stochastic differential equations , which enjoys the tractability that the first aforementioned approach has, and at the same time accounts for fluctuations, by modeling noise in terms of Brownian motion. This paper develops the latter approach to incorporate the stochasticity in the detection process to the model equations. The resulting neutron count distributions are explicitly computable. As an application of our approach we present a straightforward derivation of the well-known Feynman-Y formula. We then propose an alternative to the traditional sampling scheme of this formula, based on mean absolute deviation , known from the statistics literature to be more robust than the mean square deviation estimator. The study focuses on a single energy point model and neglects the effect of the delayed neutrons. Extensions of the approach to multiple energy levels and the incorporation of delayed neutrons are discussed, as well as further applications of the approach and its advantages over existing diffusion scale approximations. [ABSTRACT FROM AUTHOR]

The iterated fission probability (IFP) method and the Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance CHaracterization (CLUTCH) method have been implemented in several Monte Carlo codes to perform sensitivity calculations. In this work the unified adjoint and generalized adjoint calculation framework combining the IFP method and the CLUTCH method is proposed, and corresponding capabilities for calculating nuclear data sensitivity to the effective multipilication factor and reaction rate ratios are implemented in OpenMC. Results are presented on Godiva, Jezebel and a PWR fuel pin system which demonstrate that OpenMC results agree well with reference direct perturbation sensitivity coefficients. In addition, two main problems of the IFP method related to the number of latent generations, namely convergence and statistical fluctuations are investigated theoretically and numerically. This paper further demonstrates that the fission matrix approach is better suited to generate the adjoint and generalized adjoint source distribution in the CLUTCH method to overcome the drawbacks of the IFP method. [ABSTRACT FROM AUTHOR]

Nuclear data improvement, validation and completion are one of the major concerns for ensuring safety-by-design standards and economic optimization for modern nuclear facilities. Uncertainty and sensitivity analyses must be carried out to estimate the nuclear data influence on integral reactor parameters, highlighting margins of improvement needed to get full benefit from modern advanced modeling. The quality of any uncertainty quantification analysis strongly depends on the covariance matrices made available in the evaluated nuclear data libraries. In modern nuclear data libraries (e.g. JEFF, ENDF/B or JENDL) no correlations for fission yields (FY) are provided, so in the present work we propose a covariance generation methodology which yielded reasonable results for application-relevant thermal and fast neutron-induced fission product yields. The main goal was not to perform a new evaluation but to reproduce coherently the JEFF-3.1.1 fission yield library and add covariance information. We did so using the Generalized Least Square Method (GLSM) and the Marginalization techniques implemented in the CONRAD code, developed at CEA-Cadarache. Results on the thermal neutron-induced fission of 235 U , 239 Pu , 241 Pu and on the fast-neutron induced fission of 238 U , 239 Pu and 240 Pu are presented and tested on elementary fission decay heat uncertainty quantifications. [ABSTRACT FROM AUTHOR]

The FP7 EURATOM project FREYA has been executed between 2011 and 2016 with the aim of supporting the design of fast lead-cooled reactor systems such as MYRRHA and ALFRED. During the project, a number of critical experiments were conducted in the VENUS-F facility located at SCK·CEN, Mol, Belgium. The Monte Carlo code Serpent was one of the codes applied for the characterization of the critical VENUS-F cores. Four critical configurations were modeled with Serpent, namely the reference critical core, the clean MYRRHA mock-up, the full MYRRHA mock-up, and the critical core with the ALFRED island. This paper briefly presents the VENUS-F facility, provides a detailed description of the aforementioned critical VENUS-F cores, and compares the numerical results calculated by Serpent to the available experimental data. The compared parameters include keff, point kinetics parameters, fission rate ratios of important actinides to that of U235 (spectral indices), axial and radial distribution of fission rates, and lead void reactivity effect. The reported results show generally good agreement between the calculated and experimental values. Nevertheless, the paper also reveals some noteworthy issues requiring further attention. This includes the systematic overprediction of reactivity and systematic underestimation of the U238 to U235 fission rate ratio. [ABSTRACT FROM AUTHOR]

The so-called Iterated Fission Probability (IFP) method has provided a major breakthrough for the calculation of the adjoint flux and more generally of adjoint-weighted scores in Monte Carlo eigenvalue calculations. So far, IFP has been exclusively devoted to the analysis of the standard k -eigenvalue equation, by resorting to a formal identification between the adjoint fundamental eigenmode φ k † and the neutron importance I k . In this work, we extend the IFP method to the α -eigenvalue equation, enabling the calculation of the adjoint fundamental eigenmode φ α † and the associated adjoint-weighted scores, including kinetics parameters. Such generalized IFP method is first verified in a simple two-group infinite medium transport problem, which admits analytical solutions. Then, α -adjoint-weighted kinetics parameters are computed for a few reactor configurations by resorting to the Monte Carlo code Tripoli-4 ®, and compared to the k -adjoint-weighted kinetics parameters obtained by the standard IFP. The algorithms that we have developed might be of interest in the interpretation of reactivity measurements, in the context of reactor period calculations by Monte Carlo simulation. [ABSTRACT FROM AUTHOR]

The nuclear data of n+ 239 Pu reaction for the incident energy up to 200 MeV are currently required for the study of new-generation nuclear systems. To meet this requirement, all cross sections, angular distributions, energy spectra, double differential cross sections of neutron, proton, deuteron, triton, helium-3 and alpha emissions and the number of neutron per fission for n+ 239 Pu reaction are consistently calculated and analyzed by theoretical nuclear models in the energy range of E n ⩽ 200 MeV . The theoretical models include the optical model, the distorted wave Born approximation theory, the Hauser-Feshbach theory, the fission model, the evaporation model, the exciton model and the intranuclear cascade model. The calculated results are compared with recent experimental data and the evaluated data from ENDF/B-VII and JENDL-4. [ABSTRACT FROM AUTHOR]

The laser-Compton scattering (LCS) phenomenon is one known method for producing energetic and high-intensity gamma rays with variable photon spectrums. Availability of high-intensity gamma rays has made probing the (γ,n)-based photonuclear transmutation possible and worth consideration. Parametric studies of LCS have shown that the gamma ray’s spectrum is tunable and dependent upon various facility-related parameters. In this paper, a systematic optimization study of the LCS spectrum has been performed in view of the incident laser energy, electron beam energy and collimation angle. This optimization has been done with the aim to maximize the transmutation reaction rates of long-living fission products (LLFPs) such as 129 I, 135 Cs, and 137 Cs that have an isotopic composition of typical LWR spent fuel. In addition, the (γ,n) and (γ,2n)-reaction rates have been simultaneously calculated in order to evaluate the feasibility of photonuclear transmutation using the (γ,n)-reaction. It has been shown that careful optimization of the LCS spectrum noticeably enhances the (γ,n)-reaction rate without requiring any isotopic separation of the iodine and cesium targets. Furthermore, in order to enhance the energy efficiency of the LCS-based photo-transmutation system, a novel multiple laser-Compton scattering extraction (MULEX) concept has been introduced in this work, and it has been shown that the total production of the LCS photons can easily be increased by a factor of 10 or 20 in a single accelerator system. [ABSTRACT FROM AUTHOR]

A hybrid fission-fusion reactor with a Dense Plasma Focus (DPF) as a fusion core and the dual layer fissionable blanket as the energy multiplier were conceptually designed. A cylindrical DPF, energized by a 200 kJ bank energy, is considered to produce fusion neutron, and these neutrons drive the subcritical fission in the surrounding blankets. The emphasis has been placed on the safety and energy production with considering technical and economical limitations. Therefore, the k eff-t of the dual cylindrical blanket was defined and mathematically, specified. By applying the safety criterion ( k eff-t ≤ 0.95), the geometrical and material parameters of the blanket optimizing the energy amplification were obtained. Finally, MCNPX code has been used to determine the detailed dimensions of the blankets and fuel rods. [ABSTRACT FROM AUTHOR]

DeCART multi-group cross section library generation procedure based on the KAERI library generation system was improved using the new alternative procedures based on the Monte Carlo slowing down calculations. In the new alternative procedures, the resonance integral tables were generated by solving the slowing down equation with continuous energy Monte Carlo calculations and the new procedures for the library generation were incorporated for the additional improvements. Most of fission product and actinide nuclides are processed as resonant nuclides to reduce the errors which are resulted from the approximate treatment of their resonance cross sections, and the one-group effective FPYs (Fission Product Yield) which are calculated by fission reaction rates are incorporated for improving the accuracy of the burnup calculations. Using the new procedure and ENDF/B-VII.1 evaluated nuclear data library, new DeCART cross section libraries are generated and they are tested for various benchmark calculations for pin-cell, FA, and core depletion problem against the McCARD reference solutions. The results confirmed that the new generated libraries can be used in the PWR SMR (Small Modular Reactor) core design. [ABSTRACT FROM AUTHOR]

Composite delayed energy spectra from epithermal neutron induced fission of 235 U have been measured for twelve delay-time intervals extending from 0.12 to 150 s. These data provide a comprehensive set of composite spectra with enhanced sensitivity to neutrons emitted in less than 1 s following fission. Extensive comparisons between these measurements and the composite spectra obtained by summation technique are presented and discussed herein. [ABSTRACT FROM AUTHOR]