FNG HCPB Tritium Breeder Module Mock-Up Benchmarking of OpenMC and Uncertainty Quantification

To assure tritium self-sufficiency in future fusion reactors such as DEMO the accuracy of TRP calculations has to be demonstrated within the design uncertainties. A new neutronics experiment representing a mock-up of the Water Cooled Lithium Lead (WCLL) Test Blanket Module (TBM) is under preparation at the Frascati neutron generator (FNG) with the objective to provide an experimental validation of accuracy of nuclear data and neutron transport codes for the tritium production rate (TPR) calculations. The mock-up will consist of LiPb bricks, EUROFER plates and Perspex substituting water. The mock-up will be irradiated by 14 MeV neutrons at the FNG facility, and the TPR and detector reaction rates will be measured using Li2CO3 pellets and activation foils placed at different positions up to about 55 cm inside the mock-up. Computational pre-analyses for the design of the WCLL neutronics experiment using the SUSD3D sensitivity/uncertainty (S/U) code system is described and compared with the results of some similar FNG experiments performed in the past, in particular the FNG HCPB Tritium Breeder Module Mock-up (2005) and FNG-HCLL Tritium Breeder Module Mock-up (2009). The objective of the pre-analysis is to provide the calculated nuclear responses including the uncertainties due to the uncertainties in nuclear data and thus contributes to the optimisation of the design of the experimental set-up.

ASUSD nuclear data sensitivity and uncertainty program package: Validation on fusion and fission benchmark experiments

Numerical benchmark of SuperMC3.2 with HCPB mock-up experiment

Analyses of radiation fields resulting from a deuterium-tritium (DT) plasma in fusion devices is a critical input to the design and validation of many aspects of the reactor design, including, shielding, material lifetime and remote maintenance requirements/scheduling. Neutronics studies, which perform in-depth analysis are typically performed using radiation transport codes such as MCNP, TRIPOLI, Serpent, FLUKA and OpenMC. The Serpent 2 Monte-Carlo code, developed by VTT in Finland, is the focus of this work which seeks to benchmark the code for fusion applications. The application of Serpent 2 in fusion specific analysis requires validation of the codes performance in an energy range, and a geometrical description, which significantly differs to conventional nuclear fission analysis, for which the code was originally developed. A Serpent model of the Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) mock up experiment has been prepared and the calculated results compared against experimental data, as well as the reference Monte Carlo code MCNP. The analysis is extended to a model of DEMO with HCPB blanket concept. For this model, the flux, nuclear heating, tritium production and DPA are calculated, all of which are integral nuclear responses in fusion reactor analysis. In general, a very good agreement is demonstrated for both of the benchmarks, with any discrepancies pinpointed to different physics models implemented.

Neutronics experiment on a helium cooled pebble bed (HCPB) breeder blanket mock-up

An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL) concept will be performed in 2008 in the Frascati Neutron Generator (FNG) in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n) and (n,3n) reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

Sensitivity and uncertainty analyses of the tritium production in the HCPB breeder blanket mock-up experiment

Numerical benchmarks TRIPOLI − MCNP with use of MCAM on FNG ITER bulk shield and FNG HCLL TBM mock-up experiments

Released in 2009, the Serpent Monte Carlo code has established itself as a highly efficient and powerful simulation code for nuclear systems analysis. Originally developed for reactor physics applications, the scope of the code now extends to coupled multi-physics simulations and photon transport. The latter has allowed adoption of the code by the fusion neutronics community following developments of a coupled neutron-photon capability in 2014 and the ability to handle complex geometry types in 2016. The code is well validated for the energy regimes and geometry types one can expect in fission reactor analysis. Over the course of recent years a benchmarking effort has been undertaken for application of the code to nuclear fusion. Compared to nuclear fission, or accelerator based applications, the underlying particle interaction phenomena differ greatly at the energies expected in a fusion reactor as well as the specific responses that are of interest. In this paper, a novel weight window generation implementation in Serpent is investigated. The applicability of this method is demonstrated for the Frascati Neutron Generator (FNG) bulk blanket and shield experiment, part of the SINBAD database, and a DEMO helium cooled pebble bed (HCPB) computational model. A comparison is performed against MCNP using weight windows generated with ADVANTG. Excellent agreement is found for the specified tallies and the significant efficiency gain using weight windows generated using both methods is comparable. A robust variance reduction method implementation is fundamental to applications to fusion neutronics and as such, this work is an important step in deployment of Serpent for this type of analysis.

Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries

Neutronics experiments on HCPB and HCLL TBM mock-ups in preparation of nuclear measurements in ITER

Predictions of material activity in commercial fusion conditions predominantly rely on computational methods, due to a lack of data on long-term effects of high-energy neutron irradiation on structural steels. Consequently, this could result in a bias due to uncertainties in nuclear data used. This work focused on modelling neutron activation of four structural steels in a fusion reactor environment after 20 years of operation. Eurofer, F82H and G91, were assessed as candidate in-vessel materials, whereas SS316L(N)-IG was solely modelled in the vacuum vessel. Activation calculations were performed using the inventory code FISPACT-II using inputs from Monte-Carlo transport simulations performed with OpenMC. The study employed a one-dimensional reactor model with a Helium-Cooled Pebble Bed (HCPB) tritium-breeding blanket design. With the XSUN-2022 code package, a nuclear data sensitivity and uncertainty analysis on production cross-sections of relevant radio-nuclides was carried out. Eurofer and F82H steels exhibited significantly higher resistance to neutron activation than G91 and SS316L(N)-IG. At 100 years after shutdown, none of the steels reached UK low-level waste (LLW) activity levels in the first wall. In the rear of the back-support structure (BSS) of the reactor blanket, all assessed steels reached LLW levels within approximately 30 to 45 years of reactor shutdown. It was found that the vacuum vessel (SS316L(N)-IG) would not be classifiable as LLW for several centuries. Dominant radio-nuclides for each material were identified with FISPACT-II to carry out the uncertainty analyses. The calculated uncertainties were too small to affect the waste disposal options for the first wall within 100 years, but the time-to-reach LLW for BSS and vacuum vessel steel could be uncertain by up to approximately 3 and 6 years, respectively.

The prediction capability for tritium production and other reaction rates in various systems configurations for a series of the USDOE/JAERI collaborative fusion blanket experiments

The propagation of nuclear data uncertainties in fusion neutronics calculations is presented in this paper. The uncertainty propagation employs the random samples of neutron cross sections and secondary particle energy/angular distributions generated by the SANDY code as nuclear data in the transport simulation of the Monte Carlo (MC) code OpenMC. The random samples are obtained from stochastic sampling employing covariances in nuclear data libraries. In this work, uncertainties in nuclear data result in perturbed neutron flux distributions that are then propagated to the gamma heating and tritium production rates in the Fusion Neutron Source clean benchmark experiments on vanadium, beryllium, tungsten, iron, copper, and graphite assemblies, which were irradiated with a 14-MeV deuterium-tritium neutron source from the Shielding Integral Benchmark Archive and Database (SINBAD). The uncertainty analysis results show that for the beryllium assembly, the tritium production uncertainties are dominated by the 9Be cross sections, while the cross sections of 6Li and the impurities present have an insignificant effect on the tritium production. In addition, the gamma heating in the vanadium assembly has the largest uncertainty (up to 23%, with impurities contributing less) among the materials analyzed, followed by graphite (~ 20%), tungsten (17%), iron (14%), and copper (< 6%). These results are important for the application of best estimate plus uncertainty methods, verification and validation, and design of fusion reactors and power plants.

The accurate assessment of tritium breeding parameters within fusion blankets is crucial for future magnet-confined fusion machines to realize fuel self-sufficiency. Such an assessment can be conducted using the simulation approach with nuclear data of high-fidelity and, most importantly, validated against experimental data. In this paper, we report neutronics experiment studies carried out on a mock-up of the water-cooled ceramic breeder blanket of the China fusion engineering test reactor (CFETR), irradiated at a deuterium–tritium (D–T) neutron source. The mock-up’s nuclear responses to 14 MeV neutrons, including tritium production rates (TPR) and neutron-induced reaction rates, are simulated and validated by experimental results. Redundant measurement techniques are used, including Li2TiO3 pellets for offline TPR measurements and the developed lithium glass detector with a significantly reduced size for online TPR measurements. The validation of the TPR value is complemented by the experimental evaluation of the neutron-induced reaction rates for Au and Zr foils. All experimental results are analyzed using the MCNP-4C code and FENDL-3.0 nuclear data library. The source term for the Monte Carlo simulation is built using a newly-developed method based upon the modeling of the depth profiling of tritium in the tritiated target. The experiments are in good agreement with the simulations; ratios of the calculation to experimental results (C/E) on TPR are found to be 0.97–1.08. The influence of the first wall tungsten armor on the mock-up nuclear responses is also studied.