Evgeny Ivanov - Academia.edu (original) (raw)

Papers by Evgeny Ivanov

EPJ Nuclear Sciences & Technologies

The use of Data Assimilation methodologies, known also as a data adjustment, liaises the results ... more The use of Data Assimilation methodologies, known also as a data adjustment, liaises the results of theoretical and experimental studies improving an accuracy of simulation models and giving a confidence to designers and regulation bodies. From the mathematical point of view, it approaches an optimized fit to experimental data revealing unknown causes by known consequences that would be crucial for data calibration and validation. Data assimilation adds value in a ND evaluation process, adjusting nuclear data to particular application providing so-called optimized design-oriented library, calibrating nuclear data involving IEs since all theories and differential experiments provide the only relative values, and providing an evidence-based background for validation of Nuclear data libraries substantiating the UQ process. Similarly, it valorizes experimental data and the experiments, as such involving them in a scientific turnover extracting essential information inherently contained ...

EPJ Web of Conferences, 2021

Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppl... more Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppler broadening, moderation ratio, and, sometimes, neutron gas temperature etc.) and on variations of thermal-physics parameters (temperature distributions, fuel elongation and moderator density). Usually, in a rough reactor analysis one ignores the very details of temperature distributions largely staying in a frame of so-called adiabatic assumptions (when temperature and density distribution are changing in sync keeping given spatial shapes). In majority of practical applications the radially distributed temperature fields are represented as monotonically smeared ones as if fissile and other materials are homogeneously mixed. Moreover, no one measurement technique allows counting precise correlation between reactivity feedback and in-pellet temperature and materials space-time distributions. However, if fuel is made of Mixed Oxide Plutonium-Uranium compound the behavior of Light Water Re...

International audienceA sensitivity benchmark exercise was organized within the scope of the Unce... more International audienceA sensitivity benchmark exercise was organized within the scope of the Uncertainty Analysisin Modeling (UAM) project of the OECD/Nuclear Energy Agency (NEA) to develop andcompare methods for the sensitivity and uncertainty computations of the effectivemultiplication factor (keff) and the effective delayed neutron fraction (eff). Several solutionswere received using different codes, both deterministic (SUSD3D, SNATCH) and MonteCarlo (TSUNAMI-3D, XSUSA, SERPENT2, MCNP6). In this paper the performances ofseveral codes and methods for the keff sensitivity and uncertainty computations are intercompared. The sensitivity and uncertainty codes were applied to the SNEAK-7A and -7B fastneutron benchmark experiments from the IRPhE database. Good general agreementbetween the sensitivities, both for integral values and sensitivity profiles, was observed

International audienceBest Estimate Plus Uncertainty (BEPU) in its basis pretends on an exact use... more International audienceBest Estimate Plus Uncertainty (BEPU) in its basis pretends on an exact use of high-fidelity simulations in a process of safety assessment. Despite on shared definition of BEPU has not been yet presented it seems to be a Decision Making (DM) support tool. BEPU needs, of course, in a consistent experiment-based UQ otherwise it could make a little sense for characterization of safety margins and, in its order, in safety assessment. Panel discussion stresses the needs to provide consistent system of criteria intended to prioritize existing IEs with respect to their impacts on VetUQ process. Given that IEs bring unique objective information inaccessible via differential experimentation one can see that to make this information useful IEs should be evaluated quantifying total uncertainties of each single IE and correlations between IE cases

EPJ Web of Conferences, 2021

The paper presents our first results of the exercise III-I-2c from the OECD-NEA UAM-LWR benchmark... more The paper presents our first results of the exercise III-I-2c from the OECD-NEA UAM-LWR benchmark intended to an elaboration of the methodology of uncertainty propagation. The considered case studied a full PWR core behavior in fast (~0.1 sec) rod ejection transient. According to the benchmark, the core represented a Hot Zero Power state. Authors used brute-force sampling propagating nuclear data and thermo-fluid uncertainties using 3D computational IRSN chain HEMERA. It couples the reactor physics code CRONOS and thermal-hydraulic core code FLICA4. The nuclear data uncertainties were represented in a form of cross sections standard deviations (in percentage of the mean cross sections values) supplied by the UAM team. In addition to the original benchmark, the study includes a case with an increased power peak by supplementary rod ejection, i.e. with higher reactivity. Both the results are similar to what we obtained in the mini-core rod ejection: the power standard deviation follow...

Annals of Nuclear Energy, 2021

Abstract The paper discusses some topics related to a validation of multi-physics modeling. Since... more Abstract The paper discusses some topics related to a validation of multi-physics modeling. Since validation belongs to a category of decision-making processes, it should be prepared by dedicated scientific researches. In particular, the validation means a kind of characterization of the predictive capability maturity of given tools, libraries and calculational models. The compliance criteria might be expressed in terms of achievable accuracy or, inverse, in terms of somehow identified and quantified uncertainties. Despite the uncertainties, as such, might not be measured or compared with something measurable all the judgments should rely on reality, i.e., be supported by an evidence-based background. In practice it does require to eliminate subjective statements, if any, replacing them with something inferred from objective observations, including representative integral experiments. Unfortunately, in many fields like, among others, multi-physics simulations, we have not statistically sufficient number of high-fidelity and confident experiment-based benchmarks. In addition, because of technological and safety constraints, what is needed lies, largely, beyond the experimental domain. This is why, assessors have to rely on numerous, but partially representative experiments. These data could be treated using one or other Data Assimilation techniques to provide correction factors and uncertainties to single- and few-physics modules all having an evidence-based background. Then, coupling these pre-validated modules and their uncertainties, we could estimate uncertainties (and accuracies) in an application domain using one of wide range error propagation techniques. Thus, combining experiments-grounded and calibrated uncertainties, we are providing consistent, evidence-based background for validation. The last phase – constrained uncertainty propagation – was illustrated with an example of one international standard problem on transient initiated in LWR core by inadvertent Control Rod ejection.

Annals of Nuclear Energy, 2020

This paper presents the execution and the results of a rod ejection benchmark inspired by the UAM... more This paper presents the execution and the results of a rod ejection benchmark inspired by the UAM benchmarks II-2b: cross sections uncertainties propagation during a rod ejection transient in a mini-core (3x3 PWR-type assemblies), without feedback. Performing transient computations in two groups time-dependent neutron diffusion approximation while starting from a steady-state power of 0.1409 MW, we obtained a mean power peak around 0.25 MW with a standard deviation of, nearly, 1.5%. The timedependent standard deviation curve follows the time-dependent mean power curve and the relative standard deviation, in a percentage of the mean power, increases with the mean power peak value. We found, too, that the first group diffusion coefficient and the first group absorption cross section are, practically, the only contributors to the total power peak variance. Drawbacks and findings from the exercise provide essential input for uncertainty propagation in a further multi-physics simulation of a postulated rod ejection accident.

Nuclear Technology, 2019

As a technical support organization for French Public Authorities, the Institute for Radioprotect... more As a technical support organization for French Public Authorities, the Institute for Radioprotection and Nuclear Safety (IRSN) shall participate in policy making on nuclear safety and radiation protection. In the framework of its current activity, the IRSN performs analysis and comparison of different safety assessment approaches, including Best Estimate Plus Uncertainty (BEPU), among others. Although BEPU is not yet recognized worldwide, the authors consider it a promising way to stimulate the involvement of a new generation of numerical tools and high-fidelity experiment data for nuclear safety assessment and development of nuclear technology. This paper discusses enabling factors and constraints for BEPU from technical and global points of view, relying on scientific background as well as current practice with respect to its potential role in regulation and, more generally, in a decision-making process.

EPJ Nuclear Sciences & Technologies, 2016

A resolved resonance evaluation was performed for 16 O in the energy range 0 eV to 6 MeV using th... more A resolved resonance evaluation was performed for 16 O in the energy range 0 eV to 6 MeV using the computer code SAMMY resulting in a set of resonance parameters (RPs) that describes well the experimental data used in the evaluation. A RP covariance matrix (RPC) was also generated. The RP were converted to the evaluated nuclear data file format using the R-Matrix Limited format and the compact format was used to represent the RPC. In contrast to the customary use of RP, which are frequently intended for the generation of total, capture, and scattering cross sections only, the present RP evaluation permits the computation of angle dependent cross sections. Furthermore, the RPs are capable of representing the (n, a) cross section from the energy threshold (2.354 MeV) of the (n, a) reaction to 6 MeV. The intent of this paper is to describe the procedures used in the evaluation of the RP and RPC, the use of the RPC in benchmark calculations and to assess the impact of the 16 O nuclear data uncertainties in the calculate dk eff for critical benchmark experiments.

EPJ Web of Conferences, 2016

A new 235 U capture cross-section evaluation, evaluated by ORNL and the CEA Bruyères-le-Châtel (B... more A new 235 U capture cross-section evaluation, evaluated by ORNL and the CEA Bruyères-le-Châtel (BRC) has been proposed within the CIELO project. IRSN, who participates in the CIELO project, contributes with data testing and has carried out benchmark calculations using few benchmarks, extracted from the ICSBEP database, for testing the new 235 U evaluation. The benchmarks have been selected by privileging the experiments showing small experimental uncertainties and a significant sensitivity to 235 U capture cross-section. The keff calculations were performed with both the MCNP 6 code and the 5.C.1 release of the MORET 5 code, using the ENDF/B-VII.1 library for all isotopes except 235 U, for which both the ENDF/B-VII.1 and the new 235 U evaluation was used. The benchmark selection allowed highlighting a significant effect on keff of the new 235 U capture cross-section. The results of this data testing, provided as input for the evaluators, are presented here.

Nuclear Data Sheets, 2014

Nuclear Science and Engineering, 2014

ABSTRACT The effective delayed neutron fraction b eff is of primary importance for reactivity con... more ABSTRACT The effective delayed neutron fraction b eff is of primary importance for reactivity control of fissile systems and therefore for reactor design and safety analyses. Validation of b eff calculations is complicated by the limited availability of benchmark-quality data. This paper focuses on evaluation and analysis of b eff measurements with 252 Cf-source pseudo-worth and noise methods performed at SNEAK 7A and SNEAK 7B assemblies in Germany in the 1970s. The experiments are thoroughly documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments and briefly presented in this paper. The measurements performed with the two different methods on SNEAK 7A and SNEAK 7B and other facilities systematically produce different values. Given that the noise approach presumes evolution of neutron field fluctuations in a one-point kinetic model, it was assumed that the discrepancies originate from spatial effects. A two-point kinetic model was tested to check this assumption. The paper demonstrates that the b eff measured with the noise method on the SNEAK 7A and SNEAK 7B facilities should be corrected while the 252 Cf-source pseudo-worth measurement produces an accurate value.

Nuclear Science and Engineering, 2014

ABSTRACT Criticality safety and reactor physics benchmark experiments may share certain component... more ABSTRACT Criticality safety and reactor physics benchmark experiments may share certain components of the configurations and the same experimental techniques. This results in correlations between pairs of benchmark experiments. The correlations may impact validation results of criticality safety or reactor physics calculations as well as nuclear data validation. This paper presents how experimental correlations have been established for some fast and thermal neutron spectra configurations, documented in International Handbook of Evaluated Criticality Safety Benchmark Experiments and in International Handbook of Evaluated Reactor Physics Benchmark Experiments. It also discusses the impact of those correlations on validation of the most recent nuclear data libraries.

Nuclear Data Sheets, 2014

OECD/NEA WPEC Sg. 33 Benchmark ▌ Explicitly presented by M. Salvatores (Plenary Session, March 5)... more OECD/NEA WPEC Sg. 33 Benchmark ▌ Explicitly presented by M. Salvatores (Plenary Session, March 5): "WPEC SG33 on: Methods and Issues for the Combined Use of Integral Experiments" ▌ Objective: "to study methods and issues of the combined use of integral experiments and covariance data, with the objective of recommending a set of best and consistent practices in order to improve evaluated nuclear data files".

The purpose of this document is to provide an updated overview of specific safety and radiologica... more The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radio...

Annals of Nuclear Energy, 2011

The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized... more The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR) has been investigated. If increasing quantities of 238 U, the fertile isotope in standard UO 2 fuel, are replaced by 232 Th, then a greater yield of new fissile material (233 U) is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 10 3 years. Two innovative fuel types named S90 and S20, ThO 2 mixed with 90% and 20% enriched UO 2 respectively, are compared as an alternative to standard uranium oxide (UOX) and uranium/plutonium mixed oxide (MOX) fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.

EPJ Nuclear Sciences & Technologies

The use of Data Assimilation methodologies, known also as a data adjustment, liaises the results ... more The use of Data Assimilation methodologies, known also as a data adjustment, liaises the results of theoretical and experimental studies improving an accuracy of simulation models and giving a confidence to designers and regulation bodies. From the mathematical point of view, it approaches an optimized fit to experimental data revealing unknown causes by known consequences that would be crucial for data calibration and validation. Data assimilation adds value in a ND evaluation process, adjusting nuclear data to particular application providing so-called optimized design-oriented library, calibrating nuclear data involving IEs since all theories and differential experiments provide the only relative values, and providing an evidence-based background for validation of Nuclear data libraries substantiating the UQ process. Similarly, it valorizes experimental data and the experiments, as such involving them in a scientific turnover extracting essential information inherently contained ...

EPJ Web of Conferences, 2021

Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppl... more Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppler broadening, moderation ratio, and, sometimes, neutron gas temperature etc.) and on variations of thermal-physics parameters (temperature distributions, fuel elongation and moderator density). Usually, in a rough reactor analysis one ignores the very details of temperature distributions largely staying in a frame of so-called adiabatic assumptions (when temperature and density distribution are changing in sync keeping given spatial shapes). In majority of practical applications the radially distributed temperature fields are represented as monotonically smeared ones as if fissile and other materials are homogeneously mixed. Moreover, no one measurement technique allows counting precise correlation between reactivity feedback and in-pellet temperature and materials space-time distributions. However, if fuel is made of Mixed Oxide Plutonium-Uranium compound the behavior of Light Water Re...

International audienceA sensitivity benchmark exercise was organized within the scope of the Unce... more International audienceA sensitivity benchmark exercise was organized within the scope of the Uncertainty Analysisin Modeling (UAM) project of the OECD/Nuclear Energy Agency (NEA) to develop andcompare methods for the sensitivity and uncertainty computations of the effectivemultiplication factor (keff) and the effective delayed neutron fraction (eff). Several solutionswere received using different codes, both deterministic (SUSD3D, SNATCH) and MonteCarlo (TSUNAMI-3D, XSUSA, SERPENT2, MCNP6). In this paper the performances ofseveral codes and methods for the keff sensitivity and uncertainty computations are intercompared. The sensitivity and uncertainty codes were applied to the SNEAK-7A and -7B fastneutron benchmark experiments from the IRPhE database. Good general agreementbetween the sensitivities, both for integral values and sensitivity profiles, was observed

International audienceBest Estimate Plus Uncertainty (BEPU) in its basis pretends on an exact use... more International audienceBest Estimate Plus Uncertainty (BEPU) in its basis pretends on an exact use of high-fidelity simulations in a process of safety assessment. Despite on shared definition of BEPU has not been yet presented it seems to be a Decision Making (DM) support tool. BEPU needs, of course, in a consistent experiment-based UQ otherwise it could make a little sense for characterization of safety margins and, in its order, in safety assessment. Panel discussion stresses the needs to provide consistent system of criteria intended to prioritize existing IEs with respect to their impacts on VetUQ process. Given that IEs bring unique objective information inaccessible via differential experimentation one can see that to make this information useful IEs should be evaluated quantifying total uncertainties of each single IE and correlations between IE cases

EPJ Web of Conferences, 2021

The paper presents our first results of the exercise III-I-2c from the OECD-NEA UAM-LWR benchmark... more The paper presents our first results of the exercise III-I-2c from the OECD-NEA UAM-LWR benchmark intended to an elaboration of the methodology of uncertainty propagation. The considered case studied a full PWR core behavior in fast (~0.1 sec) rod ejection transient. According to the benchmark, the core represented a Hot Zero Power state. Authors used brute-force sampling propagating nuclear data and thermo-fluid uncertainties using 3D computational IRSN chain HEMERA. It couples the reactor physics code CRONOS and thermal-hydraulic core code FLICA4. The nuclear data uncertainties were represented in a form of cross sections standard deviations (in percentage of the mean cross sections values) supplied by the UAM team. In addition to the original benchmark, the study includes a case with an increased power peak by supplementary rod ejection, i.e. with higher reactivity. Both the results are similar to what we obtained in the mini-core rod ejection: the power standard deviation follow...

Annals of Nuclear Energy, 2021

Abstract The paper discusses some topics related to a validation of multi-physics modeling. Since... more Abstract The paper discusses some topics related to a validation of multi-physics modeling. Since validation belongs to a category of decision-making processes, it should be prepared by dedicated scientific researches. In particular, the validation means a kind of characterization of the predictive capability maturity of given tools, libraries and calculational models. The compliance criteria might be expressed in terms of achievable accuracy or, inverse, in terms of somehow identified and quantified uncertainties. Despite the uncertainties, as such, might not be measured or compared with something measurable all the judgments should rely on reality, i.e., be supported by an evidence-based background. In practice it does require to eliminate subjective statements, if any, replacing them with something inferred from objective observations, including representative integral experiments. Unfortunately, in many fields like, among others, multi-physics simulations, we have not statistically sufficient number of high-fidelity and confident experiment-based benchmarks. In addition, because of technological and safety constraints, what is needed lies, largely, beyond the experimental domain. This is why, assessors have to rely on numerous, but partially representative experiments. These data could be treated using one or other Data Assimilation techniques to provide correction factors and uncertainties to single- and few-physics modules all having an evidence-based background. Then, coupling these pre-validated modules and their uncertainties, we could estimate uncertainties (and accuracies) in an application domain using one of wide range error propagation techniques. Thus, combining experiments-grounded and calibrated uncertainties, we are providing consistent, evidence-based background for validation. The last phase – constrained uncertainty propagation – was illustrated with an example of one international standard problem on transient initiated in LWR core by inadvertent Control Rod ejection.

Annals of Nuclear Energy, 2020

This paper presents the execution and the results of a rod ejection benchmark inspired by the UAM... more This paper presents the execution and the results of a rod ejection benchmark inspired by the UAM benchmarks II-2b: cross sections uncertainties propagation during a rod ejection transient in a mini-core (3x3 PWR-type assemblies), without feedback. Performing transient computations in two groups time-dependent neutron diffusion approximation while starting from a steady-state power of 0.1409 MW, we obtained a mean power peak around 0.25 MW with a standard deviation of, nearly, 1.5%. The timedependent standard deviation curve follows the time-dependent mean power curve and the relative standard deviation, in a percentage of the mean power, increases with the mean power peak value. We found, too, that the first group diffusion coefficient and the first group absorption cross section are, practically, the only contributors to the total power peak variance. Drawbacks and findings from the exercise provide essential input for uncertainty propagation in a further multi-physics simulation of a postulated rod ejection accident.

Nuclear Technology, 2019

As a technical support organization for French Public Authorities, the Institute for Radioprotect... more As a technical support organization for French Public Authorities, the Institute for Radioprotection and Nuclear Safety (IRSN) shall participate in policy making on nuclear safety and radiation protection. In the framework of its current activity, the IRSN performs analysis and comparison of different safety assessment approaches, including Best Estimate Plus Uncertainty (BEPU), among others. Although BEPU is not yet recognized worldwide, the authors consider it a promising way to stimulate the involvement of a new generation of numerical tools and high-fidelity experiment data for nuclear safety assessment and development of nuclear technology. This paper discusses enabling factors and constraints for BEPU from technical and global points of view, relying on scientific background as well as current practice with respect to its potential role in regulation and, more generally, in a decision-making process.

EPJ Nuclear Sciences & Technologies, 2016

A resolved resonance evaluation was performed for 16 O in the energy range 0 eV to 6 MeV using th... more A resolved resonance evaluation was performed for 16 O in the energy range 0 eV to 6 MeV using the computer code SAMMY resulting in a set of resonance parameters (RPs) that describes well the experimental data used in the evaluation. A RP covariance matrix (RPC) was also generated. The RP were converted to the evaluated nuclear data file format using the R-Matrix Limited format and the compact format was used to represent the RPC. In contrast to the customary use of RP, which are frequently intended for the generation of total, capture, and scattering cross sections only, the present RP evaluation permits the computation of angle dependent cross sections. Furthermore, the RPs are capable of representing the (n, a) cross section from the energy threshold (2.354 MeV) of the (n, a) reaction to 6 MeV. The intent of this paper is to describe the procedures used in the evaluation of the RP and RPC, the use of the RPC in benchmark calculations and to assess the impact of the 16 O nuclear data uncertainties in the calculate dk eff for critical benchmark experiments.

EPJ Web of Conferences, 2016

A new 235 U capture cross-section evaluation, evaluated by ORNL and the CEA Bruyères-le-Châtel (B... more A new 235 U capture cross-section evaluation, evaluated by ORNL and the CEA Bruyères-le-Châtel (BRC) has been proposed within the CIELO project. IRSN, who participates in the CIELO project, contributes with data testing and has carried out benchmark calculations using few benchmarks, extracted from the ICSBEP database, for testing the new 235 U evaluation. The benchmarks have been selected by privileging the experiments showing small experimental uncertainties and a significant sensitivity to 235 U capture cross-section. The keff calculations were performed with both the MCNP 6 code and the 5.C.1 release of the MORET 5 code, using the ENDF/B-VII.1 library for all isotopes except 235 U, for which both the ENDF/B-VII.1 and the new 235 U evaluation was used. The benchmark selection allowed highlighting a significant effect on keff of the new 235 U capture cross-section. The results of this data testing, provided as input for the evaluators, are presented here.

Nuclear Data Sheets, 2014

Nuclear Science and Engineering, 2014

ABSTRACT The effective delayed neutron fraction b eff is of primary importance for reactivity con... more ABSTRACT The effective delayed neutron fraction b eff is of primary importance for reactivity control of fissile systems and therefore for reactor design and safety analyses. Validation of b eff calculations is complicated by the limited availability of benchmark-quality data. This paper focuses on evaluation and analysis of b eff measurements with 252 Cf-source pseudo-worth and noise methods performed at SNEAK 7A and SNEAK 7B assemblies in Germany in the 1970s. The experiments are thoroughly documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments and briefly presented in this paper. The measurements performed with the two different methods on SNEAK 7A and SNEAK 7B and other facilities systematically produce different values. Given that the noise approach presumes evolution of neutron field fluctuations in a one-point kinetic model, it was assumed that the discrepancies originate from spatial effects. A two-point kinetic model was tested to check this assumption. The paper demonstrates that the b eff measured with the noise method on the SNEAK 7A and SNEAK 7B facilities should be corrected while the 252 Cf-source pseudo-worth measurement produces an accurate value.

Nuclear Science and Engineering, 2014

ABSTRACT Criticality safety and reactor physics benchmark experiments may share certain component... more ABSTRACT Criticality safety and reactor physics benchmark experiments may share certain components of the configurations and the same experimental techniques. This results in correlations between pairs of benchmark experiments. The correlations may impact validation results of criticality safety or reactor physics calculations as well as nuclear data validation. This paper presents how experimental correlations have been established for some fast and thermal neutron spectra configurations, documented in International Handbook of Evaluated Criticality Safety Benchmark Experiments and in International Handbook of Evaluated Reactor Physics Benchmark Experiments. It also discusses the impact of those correlations on validation of the most recent nuclear data libraries.

Nuclear Data Sheets, 2014

OECD/NEA WPEC Sg. 33 Benchmark ▌ Explicitly presented by M. Salvatores (Plenary Session, March 5)... more OECD/NEA WPEC Sg. 33 Benchmark ▌ Explicitly presented by M. Salvatores (Plenary Session, March 5): "WPEC SG33 on: Methods and Issues for the Combined Use of Integral Experiments" ▌ Objective: "to study methods and issues of the combined use of integral experiments and covariance data, with the objective of recommending a set of best and consistent practices in order to improve evaluated nuclear data files".

The purpose of this document is to provide an updated overview of specific safety and radiologica... more The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radio...

Annals of Nuclear Energy, 2011

The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized... more The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR) has been investigated. If increasing quantities of 238 U, the fertile isotope in standard UO 2 fuel, are replaced by 232 Th, then a greater yield of new fissile material (233 U) is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 10 3 years. Two innovative fuel types named S90 and S20, ThO 2 mixed with 90% and 20% enriched UO 2 respectively, are compared as an alternative to standard uranium oxide (UOX) and uranium/plutonium mixed oxide (MOX) fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.