Search

Cosmic-ray muons can be used for the non-destructive imaging of spent nuclear fuel in sealed dry storage casks. The scattering data of the muons after traversing provides information on the thereby penetrated materials. Based on these properties, we investigate and discuss the theoretical feasibility of detecting single missing fuel rods in a sealed cask for the first time. We perform simulations of a vertically standing generic cask model loaded with fuel assemblies from a pressurized water reactor and muon detectors placed above and below the cask. By analysing the scattering angles and applying a significance ratio based on the Kolmogorov-Smirnov test statistic we conclude that missing rods can be reliably identified in a reasonable measuring time period depending on their position in the assembly and cask, and on the angular acceptance criterion of the primary, incoming muons., to be published/submitted in/to European Physical Journal N

MPACT is a state-of-the-art core simulator designed to perform high-fidelity analysis using whole-core, three-dimensional, pin-resolved neutron transport calculations on modern parallel computing hardware. MPACT was originally developed to model light water reactors, and its capabilities are being extended to simulate gas-cooled, graphite-moderated cores such as Magnox reactors. To verify MPACT’s performance in this new application, the code is being formally benchmarked using representative problems. Progression problems are a series of example models that increase in complexity designed to test a code’s performance. The progression problems include both beginning-of-cycle and depletion calculations. Reference solutions for each progression problem have been generated using Serpent 2, a continuous-energy Monte Carlo reactor physics burnup calculation code. Using the neutron multiplication eigenvalue ke_ as a metric, MPACT’s performance is assessed on each of the progression problems. Initial results showed that MPACT’s multigroup cross section libraries, originally developed for pressurized water reactor problems, were not sufficient to accurately solve Magnox problems. MPACT’s improved performance on the progression problems is demonstrated using this new optimized cross section library.

The initial design life of nuclear power plant is 40 years. In 60 year life extending license application, the fatigue of component should be evaluated under the influence of the fatigue factors of the pressurized water reactor coolant environment. Because the original design used a more conservative analysis method, the result could not meet the requirement of Cumulative usage fatigue factor of RCC-M. An optimizing analysis method is studied, and as an example of application, optimizing fatigue analysis of Safety Injection Nozzle of Main Coolant Line is performed. The evaluation results show that the optimized fatigue analysis results meet the requirements of RCC-M.

Uncertainty quantification (UQ) was performed using the Consortium for the Advanced Simulation of Light Water Reactors (CASL) multiphysics core simulator VERA. Typically, only nuclear data cross sections are considered when trying to obtain uncertainty information in reactor simulation applications. In this paper, uncertainty in both nuclear cross section data and fuel manufacturing processes is analyzed using VERA. Uncertainties due to cross sections was determined by generating one thousand perturbed cross section libraries using the cross section covariance data provided in the evaluated nuclear data library. Uncertainty due to manufacturing was also determined using stochastic sampling and VERA. The use of similar stochastic sampling techniques for the same problems allows for the direct comparison of uncertainty stemming from the two sources of uncertainty. Sample size is considered due to the potentially large computational cost of stochastic sampling techniques, as is demonstrated in a full core depletion. It was found that for the Pressurized Water Reactor (PWR) pincell case at Hot Zero Power (HZP), the standard deviation in the neutron multiplication factor produced by material uncertainty was 101 pcm, while the standard deviation in the neutron multiplication factor produced by cross section uncertainty was 730 pcm. While the uncertainty in neutron multiplication factor due to cross section uncertainty is larger than uncertainty due to manufacturing uncertainties, neglecting manufacturing uncertainties may be an unacceptable oversight in certain high-precision simulation applications.

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnox reactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.

Advanced manufacturing has the potential to revitalize US manufacturing, with valuable applications in several industries, including aerospace, automotive, and construction. Some of these applications have clear-cut objectives (e.g., maintain component performance while reducing mass). Applications of advanced manufacturing of nuclear components have aimed at recapturing lost manufacturing capabilities or addressing maintenance of legacy reactor components. Through the Department of Energy, Office of Nuclear Energy, Transformational Challenge Reactor design and analysis thrust, applications of advanced manufacturing, in particular, additive manufacturing, to core design has yielded reactor designs that are free from conventional manufacturing constraints. For applications in core design, the multiphysics nature of the key core metrics (e.g., peak temperature, peak power) in addition to transient safety performance requirements provides a more complex set of objectives that requires more advanced modeling and simulation tools. Additive manufacturing provides high dimensional control and design flexibility to produce complex coolant channel shapes for improved heat transfer properties and low peak material temperatures. Additional mechanisms for improved heat transfer characteristics and temperature-controlled feedback mechanisms have also been explored and incorporated into designs. While some of these enhancements are not directly beneficial for the current operating pressurized water reactor fleet, benefits may be realized in specific reactor applications that have a more constrained design space (e.g., mass, size, material type) or design metrics (e.g., fuel utilization).

The WIMS (Winfrith Improved Multigroup Scheme) reactor physics code is actively being developed for whole core modelling of a range of Small Modular Reactor types including the Pressurized Water Reactor (PWR), High Temperature Reactor (HTR), and Liquid Metal Cooled Fast reactor (LMFR). These developments include the capability for whole core multiphysics modelling with neutronics and thermal hydraulic feedback, as well as methods to determine the power deposition from neutron and gamma heating. Flux solutions are obtained using a wide variety of deterministic methods including diffusion theory, SP3, and full transport with the method of characteristics and Sn discrete ordinates methods, as well as multi-group Monte Carlo methods. The SP3 method allows both steady state and time dependent transient solutions by solving the time dependent SP3 equations. A wide variety of nuclear data libraries are available with WIMS including data from the JEF3.3, ENDF/B-VII.0 and CENDL3.1 nuclear data evaluations. This paper presents validation of the latest version of the WIMS code, WIMS11, for PWR and HTR systems. Comparisons are made against physics data obtained from the OECD/NEA PWR Watts Bar multi-physics benchmark and the IAEA HTR-10 benchmark, as well as neutron and gamma heating experiments that took place on the NESSUS reactor at Winfrith in the United Kingdom. In each case, validation of WIMS has been obtained by comparison either against measured data, or results provided by other benchmark participants that have been obtained with alternative deterministic or Monte Carlo methods.

Rising awareness for the environment as well as concerns over the sustainability of fossil fuels has encouraged developed and developing countries to find alternative ways to enhance the thermal efficiency of current power systems. The thermal efficiency of power plants can be increased from 30 – 40 % up to 80 – 90 % through the implementation of a trigeneration system by recovering dissipated waste heat for other purposes. The trigeneration system can be defined as a technology that can produce simultaneous power, heating, and cooling energy from the same fuel source. Trigeneration System Cascade Analysis (TriGenSCA) methodology is an optimisation approach based on Pinch Analysis that has been used to establish the guidelines or the proper size of the trigeneration system. This paper proposes a modification of TriGenSCA by considering a multi-period of energy consumption to optimise the size of the utility in the centralised trigeneration system by considering the transmission and storage of energy losses in the Total Site system. There are six steps involved including data extraction, identification of time slices, Problem Table Algorithm (PTA), Multiple Utility Problem Table Algorithm (MU PTA), Total Site Problem Table Algorithm (TS PTA), and modified TriGenSCA. The methodology has been tested on the centralised nuclear trigeneration system in a Total Site System as a case study and results shown that thermal energy needed by the Pressurized Water Reactor (PWR) trigeneration system with transmission losses is 2,427 MW whereas thermal energy needed by the PWR trigeneration system without transmission losses is 2,424 MW. The TriGenSCA with consideration of transmission and storage energy losses is useful for engineers and designers to determine the exact value of energy for trigeneration plant.

The Generic Pressurized Water Reactor (GPWR) simulator has been used in the Nuclear I&C Laboratory at Khalifa University (KU) since 2013 to improve student performance in nuclear engineering that is a multidisciplinary field involving nuclear reactor physics, thermodynamics, fluid mechanics, thermal hydraulics, radiation, etc. The simulator, developed by Western Service Corporation, has been integrated as a teaching and educational tool in different Engineering Programs at KU (Mechanical and Nuclear engineering). This lab is used in an undergraduate course where students apply the knowledge taught from different courses such as nuclear systems, fuel cycle, thermal hydraulics, safety principle, and control functions through a virtual operating NPP simulator. This real-time, full scope and high fidelity simulator allows to perform different operating conditions such as plant startups, shutdowns, and load maneuvers; as well as normal and abnormal plant transients, and critical scenarios and accidents. Since its installation in the Nuclear I&C Laboratory at KU in 2013, thirty students have benefited from this learning simulator. The main skills and learning outcomes expected to be achieved by students through the using of this tool are (i) ability to describe different NPP components and understand different process occurring in different subsystems, (ii) explain and apply safety principles and protective protocols, and (iii) analyze and interpret the plant behavior during transient operations and when severe accidents happen.

For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study ? loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. ?, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today's reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally. This publication reflects IRSN's desire to keep an enduring record of its results and to share its knowledge.

Because of harsh radiated environmental conditions, it is necessary to use thermocouples (TCs) in the temperature instrumentation channels of a reactor coolant system (RCS) in an integrated pressurized water reactor vessel. Conventionally, resistance temperature detectors (RTDs) have been adopted for RCS temperature measurement.Therefore, we have conducted an analysis and review of instrumentation error factors in the measurement circuits of RTD and TC sensors to specify the influence on measurement accuracy for application of TCs instead of RTDs for RCS temperature instrumentation.From the review and analysis results, it is anticipated that a measurement accuracy deterioration would be an issue and that a drift range should be investigated for the anticipated operational temperature conditions.

Confocal chromatic microscopy is an optical technique allowing measuring displacement, thickness, and roughness with a sub-micrometric precision. Its operation principle is based on a wavelength encoding of the object position. Historically, the company STIL based in the south of France has first developed this class of sensors in the 90’s. Of course, this sensor can only operate in a sufficiently transparent medium in the used spectral domain. It presents the advantage of being contactless, which is a crucial advantage for some applications such as the fuel rod displacement measurement in a nuclear research reactor core and in particular for cladding-swelling measurements. The extreme environmental conditions encountered in such experiments i.e. high temperature, high pressure, high radiations flux, strong vibrations, surrounding turbulent flow can affect the performances of this optical system. We then need to implement mitigation techniques to optimize the sensor performance in this specific environment. Another constraint concerns the small volume available in the irradiation rig next to the rod to monitor, implying the challenge to conceive a miniaturized sensor able to operate under these constraints.

ACP100 NPP designed by CNNC (China National Nuclear Corporation) is a 125MWe, multi-purpose small modular reactor based on pressurized water reactor technology; it adopts the integrated reactor technology. Different application scenarios bring up different design requirements: some require high compactness, but others care more about a longer cycle length, and some may require a fully mature and conservative design; thus, multiple design choices need to be proposed. Also, the same and most important thing cared by all users is that, the design needs to be validated to satisfy the current nuclear safety standards, and lower cost would be always preferred. Core nuclear design is a key part of the whole NPP design. Basically, nuclear design target of ACP100 is to achieve a reasonable good balance during longer cycle length, larger discharge exposure for fuel assemblies, and maximally using the mature technologies, and of course, with sufficient reactivity control ability for safety assurance. Aiming at satisfying all these different needs maximally, a strategy of supplying multiple nuclear design choices is proposed for ACP100: choice 1. Boron-free plan, this is a compact design with no need for chemistry and volume system, no need for daily boron adjustment and relative waste storage; choice 2. Boron and rod co-controlled plan, this is similar with large commercial PWRs, with a lower power peak factor and suitable for broad location sites. Both choices load 57 units of the same type fuel assemblies CF3S (with height reduced from CF3 fuel assemblies) per cycle, and both adopt partial reload and shuffle fuel management strategy to achieve larger discharge exposure. Gd is loaded in the fuel rods in both choices to help control reactivity. Choice 1 loads much more control rod clusters than choice 2, and of course, reactivity adjustment and compensation during operation is totally different between them. Using suitable and reliable software to simulate the core, through large amount of optimization, both choices achieve a 24-month fuel cycle length; the average discharge exposure of fuel assemblies reach about 40000MWd/tU, which is competitive among SMRs, especially for boron-free ones; and sufficient reactivity control ability and safety margin is validated to fully meet the reactor safety requirements.

International audience; Because of the increase of intermittent renewable energies, load follow operations for French PWR will be crucial in the years to come. The goal of this study is to make realistic changes to the plant operations in order to optimize load-following without disrupting the plant. 6 discrete parameters were considered among the overlaps, speeds and maneuvering bands of the control rods. A simulator oriented model of the pressurized reactor based on APOLLO3® is used. It includes 3D neutronics calculations with point kinetics and a 0D model of the secondary system. The operating mode (G mode) was modeled, to account for the human operator of the power plant. Two objectives were considered so as to both minimize the volume of effluents generated during the transient, and maximize the core axial stability. The reaction of the power plant to a load follow transient varies greatly during the operating cycle, because of fuel depletion effects. Therefore, 4 burnup points are considered and the objectives are computed for each point and then reduced to two “whole cycle” objectives. A biobjective massively parallel asynchronous master worker evolutionary algorithm based on AMS-MOEA/D was implemented. It is a highly exploratory algorithm, suitable for black-box problems with an important computing time. The analysis of the performances of this algorithm shows that it is able to find a diversified Pareto front, with solutions that greatly improve the load follow operations for all burnup points compared to the standard control rod parameters.

CAP1400 power plant is one of the advanced Third-generation Pressurized Water Reactor power plant. To ensure the safety and stability of the power plant secondary loop during operation, the corrosion characteristic of CAP1400 secondary loop should been investigated. This paper set up a test facility to study the vapor-water two-phase flow corrosion status, which could be seen in the piping system after the control valve of heater drain line. The key parameters, such as pressure drop, temperature, and surface topography have been observed and measured, and the different corrosion characteristic in straight pipe, elbow, tee, and reducer also have been acquired in the tests. The test results and analysis could provide the basis for the corrosion resistance design of the drain pipelines in the CAP1400 nuclear power plant.

ANIMMA 2017 – Advancements in Nuclear Instrumentation Measurement Methods and their Applications, Liège, Belgique, 19-23 juin 2017; International audience; In the framework of the French National Research Agency program on nuclear safety and radioprotection, the ‘DIstributed Sensing for COrium Monitoring and Safety’ project aims at developing innovative instrumentation for corium monitoring in case of severe accident in a Pressurized Water nuclear Reactor. Among others, a new under-vessel instrumentation based on Self-Powered Neutron Detectors is developed using a numerical simulation toolbox, named ‘MATiSSe’. The CEA Instrumentation Sensors and Dosimetry Lab developed MATiSSe since 2010 for Self-Powered Neutron Detectors material selection and geometry design, as well as for their respective partial neutron and gamma sensitivity calculations. MATiSSe is based on a comprehensive model of neutron and gamma interactions which take place in Selfpowered neutron detector components using the MCNP6 Monte Carlo code. As member of the project consortium, the THERMOCOAX SAS Company is currently manufacturing some instrumented pole prototypes to be tested in 2017. The full severe accident monitoring equipment, including the standalone low current acquisition system, will be tested during a joined CEA-THERMOCOAX experimental campaign in some realistic irradiation conditions, in the Slovenian TRIGA Mark II research reactor

Recently, increasing demands on the reduction of fuel cycle costs have led to higher burnup fuel designs. According to the erbia-credit super high burnup fuel concept, developed by mixing low content of erbia to UO2 powder directly after reconversion process so that all fuel pins in a given fuel assembly are homogeneously doped, the present study aims to characterize, from a neutronic point of view, a 17 × 17 pressurized water reactor assembly enriched to 10.27 wt.% in 235 U with an erbia content of 1 at.% (i.e. 0.7 wt.%) by means of the deterministic neutronic code APOLLO2. For this purpose, a simplified thermal-hydraulic analysis was performed in order to evaluate the effects on fuel thermal conductivity of adding erbia to uranium oxide. The results obtained allow to conclude that an Er-doped assembly enriched to >5 wt.% in 235 U represents an advantageous solution for very long fuel cycles, and it is so suited for very high burnups.

In the framework of the extension of the lifetime of currently operating reactors as well as of issues connected to decommissioning, accurate calculations of neutron and gamma responses outside the reactor core are increasingly being sought. Recently Monte Carlo calculations have been extended to provide a deep penetration capability incorporated within the eigenvalue calculation. This allows, in principle, neutron and gamma ray responses quite far outside the fissile region to be calculated within the same source-iteration scheme employed to define the neutronic responses in the fissile zone. In this paper, the new algorithm is compared to the classic decoupled approach - an eigenvalue calculation followed by a fixed source one - with the point of decoupling chosen as the fission sites. Two contrasting sample problems are discussed: a small fast research reactor and a large GEN-III Pressurized Water Reactor. The latter problem highlights the role of superhistories in maintaining the fundamental mode. © The Authors, published by EDP Sciences, 2017.

In a first part, this paper describes the AP1000® pressurized water reactor (PWR) design and how the AP1000 plant fundamental objectives of safety, simplification and standardization were applied throughout the design of the AP1000 plant.The second part of this paper focuses on specific safety and licensing topics: the AP1000 plant robustness for extreme events that may lead to catastrophic loss of infrastructure (such as the event that occurred at Fukushima Dai-ichi in March 2011) and the AP1000 plant compliance with the safety objectives for new plants released by the Western European Nuclear Regulators’ Association (WENRA).

The ATMEA1 reactor is an evolutionary generation III+ mid-size 1100 MWe 3-loop Pressurized Water Reactor designed by ATMEA, an AREVA and Mitsubishi Heavy Industries Ltd (MHI) joint venture. Developed primarily based on the US regulations its safety options have been reviewed by the French safety authorities which positively concluded that they are compliant with the French safety regulations. They have been reviewed by the Canadian safety authorities (CNSC) as well which concluded that, overall, the ATMEA1 design intent meets the most recent CNSC regulatory design requirements and expectations for a new nuclear power plant in Canada. The ATMEA1 reactor is now considered in several countries, such as Argentina, Brazil, Turkey and Vietnam.

The Monte Carlo Continuous Energy Burnup Code was used for numerical simulations of the pressurized water reactor (PWR) with mixed U-Th fuel. The thorium fuel was introduce to the core of the Westinghouse 4-loop PWR in order to investigate the time evolutions of fissile 233U, 235U, 239Pu, 241Pu and of fertile 232Th and 238U. The core loading pattern was designed using the Whole Assembly Seed and Blanket (WASB) approach. The following calculations show the feasibility of the proposed incore fuel management patter using the available numerical tools, developed numerical model and methods.

Life extensions of nuclear power plants to 60 and potentially 80 years of operation have renewed interest in long-term material degradation. One material being considered is concrete, with a particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database and a reliable determination of relevant neutron fluence energy cutoff value are necessary to ensure reliable risk assessment for extended operation of nuclear power plants.

Equipment Failure Root Cause Analysis (ERCA) methodology is employed in this paper to investigate the root cause for charging pump’s pressure fluctuation of chemical and volume control system (RCV) in pressurized water reactor (PWR) nuclear power plant. RCA project task group has been set up at the beginning of the analysis process. The possible failure modes are listed according to the characteristics of charging pump’s actual pressure fluctuation and maintenance experience during the analysis process. And the failure modes are analysed in proper sequence by the evidence-collecting. It suggests that the gradually untightened and loosed shaft nut in service should be the root cause. And corresponding corrective actions are put forward in details.

This research is to solve the stability analysis issue of nonlinear pressurized water reactor cores. On the basis of modeling a nonlinear pressurized water reactor core using the lumped parameter method, its linearized model is achieved via the small perturbation linearization way. Linearized models of the nonlinear core at six power levels are selected as local models of this core. The T-S fuzzy idea for the core is exploited to construct the T-S fuzzy model of the nonlinear core based on the local models and the triangle membership function, which approximates the nonlinear core model within the entire range of power level. This fuzzy model as a bridge is to cater to the stability analysis of the nonlinear core after defining its stability. One stability theorem is deduced to define the nonlinear core is globally asymptotically stable in the global range of power level. Finally, the simulation work and stability analyses are separately completed. Numerical simulations show that the fuzzy model can substitute the nonlinear core model; stability analyses verify the nonlinear core is globally asymptotically stable in the global range of power level.

Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA). Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65 Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65 Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237 Np(n,f) in the cavity capsule using BUGLE-B7.

This study is devoted to the evaluation of a new innovative micro structured fuel for future pressurized water reactor. This fuel should have the potential to increase the safety margins, lowering fuel temperatures by adding a small fraction of a high conductivity second phase material, the matrix, in the oxide fuel phase. After matrix selection, two different composite fuels were studied: first, a CERCER (CERamic = UO2 -CERamic matrix made of silicon carbide, SiC) fuel type has been investigated, the advantages of a ceramic being generally its transparency to neutrons and its high melting temperature. The study then focused on CERMET fuels (CERamic-METal). Several materials have been considered such as zirconium alloys, austenitic and ferritic stainless steels and chromium based alloys. The behavior of this fuel in a standard rod has been modeled with finite element codes and was assessed for different aspects of the cycle as neutronic studies, thermal behavior in normal, incidental and accidental conditions. Feasibility of fuels has been investigated with the fabrication and characterizations of the microstructure of composite fuels with powder metallurgy and HIP processes. Finally, a global evaluation is given.

The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

L’article presente les enjeux et l’etat des connaissances concernant les possibilites de retention de materiaux issus d’un coeur de Reacteur a Eau Pressurisee en situation hypothetique d’accident grave. On discute les situations de retention par renoyage du circuit primaire, par noyage du puits de cuve (refroidissement externe de la cuve) et on decrit le dispositif de recuperation mis en place pour le reacteur EPR TM . L’ensemble se place dans un contexte d’une demarche EPS2.

Depuis les annees soixante, et afin d’accroitre leur performance en reacteur, les composants en alliages de zirconium ont beaucoup evolue : les procedes de fabrication des alliages classiques (Zircaloy 2 et Zircaloy 4) ont ete optimises, les compositions resserrees et d’autres alliages plus performants ont ete developpes. Cet article dresse le panorama de ces evolutions pour les reacteurs a eau legere.

Angra 2, the second Brazilian nuclear power plant, began the commercial operation in 2001. The plant is a pressurized water reactor (PWR) type with electrical output of about 1350 MW. In the present work, the thermal hydraulic RELAP5-3D code was used to develop a model of this reactor. The model was performed using geometrical and material data from the Angra 2 final safety analysis report (FSAR). Simulations of the reactor behavior during steady state and loss of coolant accident were performed. Results of temperature distribution within the core, inlet and outlet coolant temperatures, coolant mass flow, and other parameters have been compared with the reference data and demonstrated to be in good agreement with each other. This study demonstrates that the developed RELAP5-3D model is capable of reproducing the thermal hydraulic behavior of the Angra 2 PWR and it can contribute to the process of the plant safety analysis.

Should an important release of radionuclides occur in the environment, one would quickly have to assess radionuclide concentration in media and foodstuff, deduce the potential exposition of concerned populations to the radiations, predict the evolution of the situation and propose different scenarios of managing the contaminated zones. To achieve this, the ASTRAL code was created, which can be used by a relatively large group of people composing crisis centres or making predictive impact studies. The starting point of the assessments is the ground deposit of radionuclides: the atmospheric dispersion phase and the cloud exposure are not taken into account here. The calculations achieved concern time evolution of radionuclide concentration in compartments of the agricultural and forest food chain as well as assessment of doses due to external and internal exposure. These values are compared to regulation limits and/or intervention levels. Different simulations of contaminated zones management can be led by applying counter-measures, dispositions meant to reduce the accident's impact on environment and populations. The seventy radionuclides taken into account in the present version are some of those met in case of accidents or during the exploitation of a Pressurized Water Reactor (PWR). They relate to 37 elements: Ag, Am, Ba, Br, Cd, Ce, Cl, Cm, Co, Cr, Cs, Eu, Fe, I, Ir, La, Mn, Mo, Nb, Nd, Ni, Np, Pm, Pr, Pu, Rb, Rh, Ru, Sb, Sm, Sn, Sr, Tc, Te, Y, Zn, Zr. On a temporal point of view, the predictions extend to 3 years after the deposition. The internet technology chosen for this version offers important advantages: the spreading and maintenance of the tool are very easy, the code may be installed on a portable computer, the access to results is much quicker, the interface is very user friendly, so that no training is needed to perform calculations. As for any radioecological tool, a large number of parameters intervene in the calculations. All parameters can be modified, even though different user profiles grant different rights in handling them.

The main temperature characteristics of a pressurized water reactor are distinguished, supporting its safety and reliable operation. The special role of the uranium fuel effective temperature is emphasized and the accuracy of the analytical determination of the power effect is increased. The calculation of the temperature distribution along the radius of the fuel rod was carried out taking into account the temperature dependence of the thermal conductivity UO 2 . The design procedure was corrected for using the Finca-Ronchi dependence for the thermal conductivity of 95% density of the theoretical one.

Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years have renewed interest in long-term material degradation in NPPs. Large irreplaceable sections of most nuclear generating stations are constructed from concrete, including safety-related structures such as biological shields and containment buildings; therefore, concrete degradation is being considered with particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the currently available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database is desirable to ensure reliable risk assessment for extended operation of nuclear power plants.

In the case of total failure of the high pressure injection (HPI) system following small break loss of coolant accident (SBLOCA) in pressurized water reactor (PWR), the break size is so small that the primary system does not depressurize to the accumulator (ACC) injection pressure before the core is uncovered extensively. Therefore, steam generator (SG) secondary-side depressurization is necessary as an accident management in order to grant accumulator system actuation and core reflood. A thermal-hydraulic analysis using RELAP5/MOD3 was made on SBLOCA with HPI-failure for Oi Units 3/4 operated by Kansai Electoric Power Co., which are conventional 4 loop PWR plants. The effectiveness of SG secondary-side depressurization procedure was investigated for the real plant design and operational characteristics. The sensitivity analyses using RELAP5/MOD3.2 showed that the accident management was effective for a wide range of break sizes, various orientations and positions. The critical break can be 3 inch cold-leg bottom break.

The result of simultaneous Main Steam Line Break (MSLB) and a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) is a depressurization in the secondary and primary system because both systems are connected through the SGTR. The OECD/NEA ROSA-2 Test 5 performed in the Large Scale Test Facility (LSTF) reproduces these simultaneous breaks in a Pressurized Water Reactor (PWR). A simulation of this Test 5 was made with the thermal-hydraulic code TRACE5. Some discrepancies found, such as an underestimation of SG-A secondary pressure during the depressurization and overestimation of the primary pressure drop after the first Power Operated Relief Valve (PORV) opening can be improved increasing the nodalization of the Upper Head in the pressure vessel and meeting the actual fluid conditions of Upper Head during the transient.

This poster presents the Hodoscope calibration curves calculation with 3D Monte Carlo code MORET. The fast neutron hodoscope is a facility of the CABRI research reactor at Cadarache (FRANCE). This hodoscope is designed to measure the fuel motion during a RIA in a pressurized water reactor. The fuel motion is measured by counting fast fission neutrons emerging from the test fuel placed in an experimental loop functioning like a Pressurized Water Reactor (T=300°C and P=155 bar), at the center of the CABRI core. The detection system of the hodoscope measures a signal which is a function of the fuel motion. The calibration curves allow then to convert the signal in a fuel mass. In order to calculate these curves, we have developed a method based on a Monte Carlo calculation code.

Electricité de France (EDF) operates 58 nuclear power plant (NPP), of the Pressurized Water Reactor (PWR) type. The loading pattern (LP) optimization of these NPP is currently done by EDF expert engineers. Within this framework, EDF R&D has developed automatic optimization tools that assist the experts. The latter can resort, for instance, to a loading pattern optimization software based on ant colony algorithm. This paper presents an analysis of the search space of a few realistic loading pattern optimization problems. This analysis leads us to introduce a hybrid algorithm based on ant colony and a local search method. We then show that this new algorithm is able to generate loading patterns of good quality.

This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

In the paper we present the Monte Carlo modelling by the means of the Monte Carlo Continuous Energy Burn-up Code of the 17x17 Pressurized Water Reactor fuel assembly designed according to the Radkowsky Thorium Fuel concept. The design incorporates the UO2 seed fuel located in the centre and (Th,U)O2 blanket fuel located in the peripheries of fuel assembly. The high power seed region supplies neutrons for the low power blanket region and thus induces breeding of fissile 233 U from fertile 232 Th. The both regions are physically separated and thus this approach is also known as either the heterogonous approach or the Seed-Blanket Unit. In the numerical analysis we consider the time evolutions of infinite neutron multiplication factor, axial/radial power density profile, 233 U, 235 U and 232 Th.

The basic strategy of Russian nuclear energy is propagation of a closed fuel cycle on the basis of fast breeder and thermal reactors, as well as the solution of the spent nuclear fuel accumulation and resource problems. The three variants of multiple Pu and U recycling in Russian pressurized water reactor concept reactors on the basis of REgenerated MIXture of U, Pu oxides (REMIX) reprocessing technology are considered in this work. The REMIX fuel is fabricated from an unseparated mixture of uranium and plutonium obtained during spent fuel reprocessing with further makeup by enriched natural U or reactor grade Pu. This makes it possible to recycle several times the total amount of Pu obtained from the spent fuel. The main difference in Pu recycling is the concept of 100% or partial fuel loading of the core. The third variant is heterogeneous composition of enriched uranium and uranium–plutonium mixed oxide fuel pins in one fuel assembly. It should be noted that all fuel assemblies with Pu require the involvement of expensive technologies during manufacturing. These three variants of the full core loadings can be balanced on zero Pu accumulation in the cycle. The various physical and economical aspects of Pu and U multiple recycling in selected variants are observed in the given work.