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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.

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.

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.

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.

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

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.

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.

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.

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.