Benchmarking and validation activities within JEFF project

The challenge for any nuclear data evaluation project is to periodically release a revised, fully consistent and complete library, with all needed data and covariances, and ensure that it is robust and reliable for a variety of applications. Within an evaluation effort, benchmarking activities play an important role in validating proposed libraries. The Joint Evaluated Fission and Fusion (JEFF) Project aims to provide such a nuclear data library, and thus, requires a coherent and efficient benchmarking process. The aim of this paper is to present the activities carried out by the new JEFF Benchmarking and Validation Working Group, and to describe the role of the NEA Data Bank in this context. The paper will also review the status of preliminary benchmarking for the next JEFF-3.3 candidate cross-section files

Operating a full tungsten actively cooled tokamak: overview of WEST first phase of operation

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017–2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m−2 were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described

Reactivity effect breakdown calculations with deterministic and stochastic perturbations analysis – JEFF-3.1.1 to JEFF3.2T1 (BRC-2009) actinides application

JEFF-3.1.1 is the reference nuclear data library in CEA for the design calculations of the next nuclear power plants. The validation of the new neutronics code systems is based on this library and changes in nuclear data should be looked at closely. Some new actinides evaluation files at high energies have been proposed by CEA/Bruyères-le-Chatel in 2009 and have been integrated in JEFF3.2T1 test release. For the new release JEFF-3.2, CEA will build new evaluation files for the actinides, which should be a combination of the new evaluated data coming from BRC-2009 in the high energy range and improvements or new evaluations in the resolved and unresolved resonance range from CEA-Cadarache. To prepare the building of these new files, benchmarking the BRC-2009 library in comparison with the JEFF-3.1.1 library was very important. The crucial points to evaluate were the improvements in the continuum range and the discrepancies in the resonance range. The present work presents for a selected set of benchmarks the discrepancies in the effective multiplication factor obtained while using the JEFF-3.1.1 or JEFF-3.2T1 library with the deterministic code package ERANOS/PARIS and the stochastic code TRIPOLI-4. They have both been used to calculate cross section perturbations or other nuclear data perturbations when possible. This has permittted to identify the origin of the discrepancies in reactivity calculations. In addition, this work also shows the importance of cross section processing validation. Actually, some fast neutron spectrum calculations have led to opposite tendancies between the deterministic code package and the stochastic code. Some particular nuclear data (MT=5 in ENDF terminology) seem to be incompatible with the current MERGE or GECCO processing codes

Fine 3D neutronic characterization of a gas-cooled fast reactor based on plate-type subassemblies

International audienceCEA neutronic studies have allowed the definition of a first 2400 MWth reference gas-cooled fast reactor core using plate-type sub-assemblies, for which the main neutronic characteristics were calculated by the so-called ERANOS 'design calculation scheme' relying on several method approximations. The last stage has consisted in a new refine characterization, using the reference calculation scheme, in order to confirm the impact of the approximations of the design route. A first core lay-out taking into account control rods was proposed and the reactivity penalty due to the control rod introduction in this hexagonal core lay-out was quantified. A new adjusted core was defined with an increase of the plutonium content. This leads to a significant decrease of the breeding gain which needs to be recovered in future design evolutions in order to achieve the self breeding goal. Finally, the safety criteria associated to the control rods were calculated with a first estimation of the uncertainties. All these criteria are respected, even if the safety analysis of GFR concepts and the determination of these uncertainties should be further studied and improved

Analysis of Dogleg Duct Experiments with 14 MeV Neutron Source Using TRIPOLI-4 Monte Carlo Transport Code

International audienceTRIPOLI-4 Monte Carlo transport code, developed by CEA, has been widely used on fission reactor physics and also can be used on fusion device neutronics. In order to verify the calculation features of TRIPOLI-4 code, a simple dogleg duct model was built to simulate the 14 MeV neutron transport based on a SINBAD fusion benchmark, called Dogleg Duct Streaming Experiment. The reaction rates in the bent duct and on the back surface of the experimental assembly for 93Nb(n,2n)92mNb, 115In(n,n')115mIn and 197Au(n,et947;)198Au neutron activation dosimeters were calculated with the TRIPOLI-4 code. To improve the calculation efficiency, variance reduction techniques of TRIPOLI-4 were also performed. The calculation results showed that the variances reduction methods of the TRIPOLI-4 code are helpful and obviously decrease the calculated time and increase the convergence efficiency. The calculation reaction rates results of 11 points inside and outside of the dogleg duct assembly were taken into account. Results from the TRIPOLI-4 simulation were compared with the experimental ones obtained from the measurements of FNS facility in Japan Atomic Energy Agency (JAEA). The benchmark results show that the TRIPOLI-4 code has a good potential to calculate and estimate neutron streaming effects in fusion device design

Continuous-Energy Adjoint Flux and Perturbation Calculation using the Iterated Fission Probability Method in Monte Carlo Code TRIPOLI-4

Pile-oscillation experiments are performed in the MINERVE reactor at the CEA Cadarache to improve nuclear data accuracy. In order to precisely calculate small reactivity variations (<10 pcm) obtained in these experiments, a reference calculation need to be achieved. This calculation may be accomplished using the continuous-energy Monte Carlo code TRIPOLI-4® by using the eigenvalue difference method. This “direct” method has shown limitations in the evaluation of very small reactivity effects because it needs to reach a very small variance associated to the reactivity in both states. To answer this problem, it has been decided to implement the exact perturbation theory in TRIPOLI-4® and, consequently, to calculate a continuous-energy adjoint flux. The Iterated Fission Probability (IFP) method was chosen because it has shown great results in some other Monte Carlo codes. The IFP method uses a forward calculation to compute the adjoint flux, and consequently, it does not rely on complex code modifications but on the physical definition of the adjoint flux as a phase-space neutron importance. In the first part of this paper, the IFP method implemented in TRIPOLI-4® is described. To illustrate the effciency of the method, several adjoint fluxes are calculated and compared with their equivalent obtained by the deterministic code APOLLO-2. The new implementation can calculate angular adjoint flux. In the second part, a procedure to carry out an exact perturbation calculation is described. A single cell benchmark has been used to test the accuracy of the method, compared with the “direct” estimation of the perturbation. Once again the method based on the IFP shows good agreement for a calculation time far more inferior to the “direct” method. The main advantage of the method is that the relative accuracy of the reactivity variation does not depend on the magnitude of the variation itself, which allows us to calculate very small reactivity perturbations with high precision. Other applications of this perturbation method are presented and tested like the calculation of exact kinetic parameters (βeff, Λeff) or sensitivity parameters

Leakage-corrected fast reactor assembly calculation with Monte-Carlo code TRIPOLI4r and its validation methodology

International audienceA leakage model based on B1 Homogeneous Equations has been recently implemented in continuous-energy Monte Carlo code TRIPOLI4. This leakage model algorithm iterates between the point-wise Monte Carlo simulation and a B1 Homogeneous Equations solver till reaching a final critical state in Monte Carlo simulation. The two advantages of our leakage model compared with the others are: we use critical flux spectrum to generate the multi-group constants for solving the B1 Homogeneous Equations; the leakage coefficients calculated are considered in point-wise Monte Carlo simulation. This leakage model is validated by a pre-designed numerical experiment simulated with continuous-energy TRIPOLI4 and the obtained results are also compared with those from SERPENT leakage model and deterministic leakage model in ECCO code. Finally, our leakage-corrected multi-group constants are used in transport theory based core calculation and they give out consistent multiplicative factor and neutronic balance

ITER transfer cask: Preliminary assessment of dose rate due to dust remained in the cask

International audienceThe Remote Handling tasks scheduled during the ITER maintenance shutdown require transportation of in-vessel components and remote-handling tools from the Vacuum Vessel (VV) ports to the Hot Cell Building (HCB). These components and tools will be moved using the Cask and Plug Remote Handling System (CPRHS). During plasma operations, plasma facing components will be highly activated by neutrons and/or contaminated with tritium. After plasma operations, activated dust will be removed from the VV but some amounts will remain. Therefore, the CPRHS may be contaminated by residual activated dust due to the transportation of these components between the VV and the HCB. As the CPRHS is not shielded, residual activated dust may lead to a residual dose rate around the CPRHS. To assess the risk of external exposition in case of human intervention for maintenance purpose inside or close to the CPRHS, dose rate estimations were performed around and inside the CPRHS for several initial dust configurations with the normalized value of 1 g of residual activated dust. The results of this study constitute a dosimetric data base and may support ITER Organization in the definition of a decontamination level and maintenance plan

Generation of 238U Covariance Matrices by Using the Integral Data Assimilation Technique of the CONRAD Code

A new IAEA Coordinated Research Project (CRP) aims to test, validate and improve the IRDF library. Among the isotopes of interest, the modelisation of the 238U capture and fission cross sections represents a challenging task. A new description of the 238U neutrons induced reactions in the fast energy range is within progress in the frame of an IAEA evaluation consortium. The Nuclear Data group of Cadarache participates in this effort utilizing the 238U spectral indices measurements and Post Irradiated Experiments (PIE) carried out in the fast reactors MASURCA (CEA Cadarache) and PHENIX (CEA Marcoule). Such a collection of experimental results provides reliable integral information on the (n,γ) and (n,f) cross sections. This paper presents the Integral Data Assimilation (IDA) technique of the CONRAD code used to propagate the uncertainties of the integral data on the 238U cross sections of interest for dosimetry applications