A STUDY OF LEAKAGE-CORRECTED TWO-STEP METHOD BASED ON THE NODAL EQUIVALENCE THEORY FOR FAST REACTOR ANALYSIS

The conventional two-step method based on the generalized equivalence theory (GET) cannot be directly applied to the fast reactor analysis since the assumption of the space-energy separability is not very valid due to a relatively long neutron mean free path. This study aims to develop a leakage-corrected two-step method for the fast reactor analysis with the aid of the albedo-corrected parameterized equivalence constants (APEC) method. The critical idea of the APEC method is to correct the homogenized group constants (HGCs) including discontinuity factors (DFs) during the nodal calculation through predetermined APEC functions. The APEC functions are functionalized in terms of the normalized leakage parameters such as a current-to-flux (CFR) ratio so that they can correct the cross-sections (XSs) and discontinuity factors by reflecting the in-situ neutron leakage information of the nodal analysis. The feasibility of the APEC-corrected two-step method was investigated by solving 5-group diffusion equations for a two-dimensional sodium-cooled fast reactor with a 6-triangle finite difference method. The 5-group HGCs for fuel assemblies were determined by using a continuous-energy Monte Carl code, and the conventional assembly discontinuity factors are also introduced for each hexagonal fuel assembly. First of all, it was demonstrated that the simple FDM scheme could reproduce the reference nodal quantities with the GET. And the APEC functions are formulated using the reference solutions to evaluate the feasibility of simple APEC functional for both XSs and DFs. Then, a smaller color-set problem was defined to determine practical APEC functions for the original benchmark, and various numerical evaluations are performed in terms of the k-eff value and nodal power distribution.

DEVELOPMENT OF THE MPACT 69-GROUP LIBRARY FOR MAGNOX REACTOR ANALYSIS USING CASL VERA

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.

Passive neutron sensing of martian subsurface from onboard rovers: results from MSL/DAN and expectations from ExoMars/Adron-RM

<p>During more than 7 years, the NASA MSL Curiosity rover is successfully traversing across the Mars surface exploring Gale crater with the Dynamic Albedo of Neutron (DAN) instrument installed onboard. This year, next generation neutron spectrometer Adron-RM is ready to be launched to Mars as a payload of the ExoMars 2020 rover. The main objectives of these instruments are analogous  and consist in the assessment of Water Equivalent Hydrogen (WEH) in the shallow martian subsurface.</p><p>The hydrogen presence significantly influences the neutron leakage spectrum because of  neutron moderation and thermalization through collisions with hydrogen nuclei. As a result, the variations of neutron flux detected onboard in different energy bands correlate with subsurface hydrogen/water abundance.</p><p>In our study, we will demonstrate scientific potential and latest results of natural neutron background measurements (called as passive measurements) by DAN. We will provide assessment on average WEH content in the area of the ExoMars 2020 landing site, which could be expected from first measurements of Adron-RM.</p>

Designing Stainless Steel Reflector at VR-1 Training Reactor

Light-water reactor cores are commonly surrounded by a stainless steel and water reflector. The reflectors are improving power distribution in the core, reducing the leakage of neutrons and thus also protecting the pressurized vessel from the neutron irradiation and the following embrittlement. Contrary to the standard procedures utilized for generation of the fuel assembly data, the reflector elements require a special approach. The major difficulty with the reflectors is represented by an absence of neutron sources in the reflector elements. Some artificial neutron source simulating the realistic source of neutrons from neutron leakage from the surrounding fuel assemblies must be added in the calculation model. The reflector data in the full-core calculations have a great impact on the power distribution in the core. The research in this field is usually focused on the square geometry, and therefore the accurate data for the hexagonal geometry are lacking. Improvements in this area are needed. Training Reactor VR-1 is used for measurements related to nuclear engineering. Department of Nuclear Reactors operating this reactor at the Czech Technical University in Prague is currently designing reflector elements containing stainless steel in order to provide measurable characteristics that can be compared to calculations realized by either Monte-Carlo codes or macroscopic core simulators. This article summarizes the methodology of development of the reflector assemblies to improve their similarity with the VVER-1000 reflector. The impact of the evaluated nuclear data is assessed. Further improvements of the proposed design is necessary to reach better agreement with the neutron spectrum in VVER-1000 reactor reflectors. The influence of evaluated data on the global characteristics was found negligible.

Measurement of leakage neutron spectra with D-T neutrons and evaluated nuclear data

Benchmarking of evaluated neutron nuclear data libraries was performed for ∼14.8 MeV neutrons on the several targets, such as gallium, graphite, silicon carbide, uranium and tungsten samples. The experiments were performed at China Institute of Atomic Energy (CIAE). The neutron leakage spectra from the samples were measured at 60◦ and 120◦ by a TOF technique with a BC501A scintillation detector. The measured spectra are rather well reproduced by MCNP-4C simulations with the CENDL3.1, JENDL-4.0 and the new release ENDF/B-VIII.0, JEFF-3.3 evaluated nuclear data libraries and so on. There have some difference between experiments and simulations for the elastic and inelastic contributions in the partial energy range. And the discrepancies of the neutron leakage spectra in the MCNP simulations originate simply from the differences in the spectra distributions of the neutron reaction channels in the evaluated nuclear data libraries.

Validation of heavy water cross section using AmBe neutron source

Physical quantities derived from integral experiments can usually be measured much more accurately than that from differential nuclear data. The accurate knowledge of integral parameters provide excellent grounds for testing and tuning differential data such as, for instance, cross sections. Measurement of neutron leakage spectra with 252Cf neutron source located at sphere center is often used for integral experiments. While this type of experiments provide information for cross section tuning, however, care must be taken to avoid misleading interpretation, namely, at high energies due to the very low portion of high energy neutrons in 252Cf spectrum. This issue can be alleviated by the use of point source with different spectra shape. For that purpose one suitable candidate seems to be the AmBe neutron source which has a relatively high average energy and peak character of emitted neutrons. Indeed, AmBe seems an interesting option because the calculated leakage neutron spectra are not very sensitive to the input shape of the neutron spectra. Thus the neutron leakage spectra calculated using tabulated of International Organization for Standardization spectra is nearly the same as stilbene measured AmBe spectra as an input.

Heterogeneous effects in simulating a fast nuclear reactor on the BFS test facility

Simulating fast neutron reactor cores for comparing experimental and calculated data on the reactor neutronics characteristics is performed using zero power test stands. The BFS test facilities in operation in Russia (Obninsk) are discussed in the present paper. The geometrical arrangement of materials in the cores of the simulated reactors (fuel pins, fuel assemblies, coolant geometry) differs from the simulation assembly on the BFS. This can cause differences between the experimental results obtained at the BFS and theoretical calculations even in the case when homogenized concentrations of all materials of the reactor are thoroughly observed. The resulting differences in neutronics parameters due to the geometry of arrangement of materials with the same homogeneous concentrations are referred to as the heterogeneous effect. Heterogeneous effects tend to increase with increasing reactor power and its size, mainly due to changes in the neutron spectra. Calculations of a number of functional values were carried out for assessing the heterogeneous effects for different spatial arrangements of the reactor materials. The calculations were performed for the following cases: a) heterogeneous distribution of materials in accordance with the design of a fast reactor; b) heterogeneous arrangement of materials in accordance with the capabilities and design features of the BFS test facility; c) homogeneous representation of materials in the reactor core and breeding blankets. The configuration of materials in accordance with the design data for fast reactors of the BN-1200 type was accepted as the basic calculation option, relative to which the effect called the heterogeneous shift of the functional value (HSF) was calculated. The effect of neutron leakage on the HSF obtained as the result of calculations using different boundary conditions was estimated. All calculations were carried out for the same homogeneous concentrations of all materials for all the above three configurations. Calculations were carried out as well for the case when plutonium metal fuel was used in the BFS. The values of the following functionals were calculated for different cases of arrangement of materials: the effective multiplication factor (reactivity), the sodium void reactivity effect, the average energy of fission-inducing neutrons, and the ratios of radioactive capture cross-sections to fission cross-sections for 239Pu. The calculations were performed using the Serpent 2.1.30 (VTT, Finland) Monte Carlo software package for neutronics simulations and ENDF/B-VII.0 and JEFF-3.1.1 evaluated nuclear data libraries. The effects of various options of material arrangement on the values of keff were found to be the greatest (about 1.6%) for the case when fissile material in the form of dioxide is replaced with metal fissile material. Homogenization of the composition reduces the keff value by about 0.4%. The average energy of fission-inducing neutrons depends to a significant extent on the leakage of neutrons and the presence of sodium (the average energy of neutrons increases and reaches in the presence of sodium about 100 keV, that is, it increases by about 11–13%). Replacing fissile material metal with its dioxide in the BFS test facility (while maintaining homogeneous concentrations, including that of oxygen) allows reducing the average energy of fission-inducing neutrons by about 60 keV. The highest values of HSF, reaching 65%, are observed when calculation of sodium void reactivity effect is performed with materials distributed homogeneously; however, HSF is equal to 1.5% when calculation of the reactor mock-up assembled on the BFS is performed. In the absence of neutron leakage (infinitely extended medium), the sodium void reactivity effect becomes positive and the HSF is equal to 4–7%. The heterogeneous effect of α for 239Pu noticeably (6–8%) depends only on the replacement of metallic plutonium with its dioxide (maintaining, of course, the homogeneous concentrations).