Validating the Serpent-Ants Calculation Chain Using BEAVRS Fresh Core HZP Data

The Serpent Monte Carlo code and the Serpent-Ants two step calculation chain are used to model the hot zero power physics tests described in the BEAVRS benchmark. The predicted critical boron concentrations, control rod group worths and isothermal temperature coefficients are compared between Serpent and Serpent-Ants as well as against the experimental measurements. Furthermore, radial power distributions in the unrodded and rodded core configurations are compared between Serpent and Serpent-Ants. In addition to providing results using a best practices calculation chain, the effects of several simplifications or omissions in the group constant generation process on the results are estimated. Both the direct and two-step neutronics solutions provide results close to the measured values. Comparison between the measured data and the direct Serpent Monte Carlo solution yields RMS differences of 12.1 mg/kg, 25.1 × 10-5 and 0.67 × 10-5 K-1 for boron, control rod worths and temperature coefficients respectively. The two-step Serpent-Ants solution reaches a similar level of accuracy with RMS differences of 17.4 mg/kg, 23.6 × 10-5 and 0.29 × 10-5 K-1. The match in the radial power distribution between Serpent and Serpent-Ants was very good with the RMS and maximum for pin power errors being 1.31 % and 4.99 % respectively in the unrodded core and 1.67 %(RMS) and 8.39 % (MAX) in the rodded core.

Neutron Cross Section Generation for PWR MOX Fuel Assemblies with SCALE and Serpent

In this study, the SCALE/TRITON code (based on deterministic method) and the Serpent 2 code (based on Monte Carlo method) were utilized to prepare the group constants of the pressurized water reactor (PWR) mixed-oxide (MOX) fuel assemblies for transient analyses of PWR MOX fueled cores in normal operation and control rod ejection accident condition with 3D reactor kinetics codes. The PWR MOX fuel assemblies were modeled with TRITON and Serpent, and their infinite neutron multiplication factors (k-inf) versus burnup and respective two-group neutron cross sections were calculated and compared against the available benchmark data obtained with the HELIOS code. The comparative results generally show a good agreement between TRITON and Serpent with HELIOS within 643 pcm for the k-inf values and within 5% for the two-group neutron cross sections. Therefore, it indicates that the TRITON and Serpent models developed herein for the PWR MOX fuel assemblies can be applied to group constant generation to be further used in transient analyses of PWR MOX fueled cores.

SMALL MODULAR PWR DESIGN FOR TRU RECYCLING WITH McCARD-MASTER TWO-STEP PROCEDURE

In our previous study, a small modular PWR core was designed for TRU (Transuranics) recycling with multi-recycling scheme with a typical two-step procedure using DeCART2D/MASTER code system in which the lattice analysis for producing homogenized group constant was performed by DeCART2D while whole core analysis was conducted by MASTER code. However, the neutron spectrum hardening of the LWR core loaded with TRU requires validating the multi-group cross section library and resonance self-shielding treatment method in lattice calculation. In this study, a new procedure using McCARD/MASTER was used to analyze the SMR core, in which the lattice calculation was performed by a Monte Carlo code called McCARD with a continuous energy library to generate homogenized two-group assembly cross sections. The SMR core analysis was performed to show neutronic characteristics and TRU mass flow in the SMR core with TRU multi-recycling. The result shows that the analyses on the neutronic characteristics and TRU mass flow using the McCARD/MASTER code system showed good agreement with the previous ones using the DeCART2D/MASTER code system. The neutronic characteristics of each cycle of the core satisfied the typical limit of a commercial PWR core and the SMR core consumes effectively TRU with net TRU consumption rates of 8.46~14.33 %.

Comparison of functional outcome of ultrasound guided glenohumeral and subacromial methylprednisolone acetate injections in frozen shoulder

<p class="abstract"><strong>Background:</strong> Frozen shoulder also known as adhesive capsulitis is a common cause for limitation of motion and pain of the shoulder joint with an incidence of 2-5% in the general population. The aim of the study was to assess the functional outcome of ultrasound guided glenohumeral and subacromial methylprednisolone acetate steroid injection in patients with frozen shoulder.</p><p class="abstract"><strong>Methods:</strong> 120 patients with frozen shoulder were randomly divided into 2 groups according to site i.e. glenohumeral and subacromial. Ultrasound-guided methylprednisolone acetate (80 mg) injection was administered. At follow up pain was being measured using VAS scale and functional outcome was measured using the DASH score and Constant score at day 0, 1 followed by 1, 3, 6 weeks and at 3 months.<strong></strong></p><p class="abstract"><strong>Results:</strong> The VAS score of glenohumeral decreased from 6.5±0.95 to 2.25±0.86 at 3 months and in subacromial it decreased from 6.37±1.07 to 2.3±0.72. DASH score in glenohumeral group decreased from 66.95±9.9 to 27.45±9.31 at 3 months and in the subacromial group it decreased from 67.52±10.65 to 26.81±11.14 at 3 months. In glenohumeral group Constant score increased from 46.63±7.18 to 72.66±7.38 at 3 months and in the subacromial group it increased from 47.92±6.91 to 70.28±6.97 at 3 months.</p><p class="abstract"><strong>Conclusions:</strong> Spontaneous recovery does not necessarily occur even after a long period, so we recommend that these modalities should be offered to all patients with frozen shoulder and it would be of more value if carried out at an early stage of the disorder.</p>

Neutronic Methodological Benchmarks With Simplified Geometries for the Gas Cooled Reactor Group Constant Generating Tools

In the ESNII+ EU FP7 project, a reactor physics benchmark aiming at the whole core calculation with the reflectors and detailed description of the structural elements was specified. This benchmark is based on the 2009 CEA concept of the ALLEGRO core. Fixed nominal technological data at nominal reactor state (geometry, composition) were prescribed which had to be modified in specified calculation branches according to different types of the thermal expansion and control rod positions. The parameters of the point kinetic model to be applied in a system thermal hydraulic code had to be determined this way. Static mechanical models of the expansion processes were specified by the benchmark. The goal of the calculation exercise was to verify the reactor physics codes, namely to get information about the modelling uncertainties and — after — their influence on the calculated results of the safety analyses. The obtained deviations between the participants are characterizing the user effects, the modelling uncertainties and the influence of the nuclear data differences all, without the possibility of their separation because of the complexity of the benchmark problem. A conclusion could be drawn that a step by step procedure starting from simple problems (homogenous material, Wigner-Seitz cell or subassembly in asymptotic approach) is necessary if we wish to identify the reasons of the deviations. For the Doppler effect, a decision was made in this direction already in the ESNII+ project where an infinite regular lattice problem without any leakage had to be solved. This approach of the simplicity is followed by the present benchmarks (one rod and one assembly), but extending the simple benchmarks with burnup calculations and taking into account leakage in asymptotic approximation by neglecting the complicated processes necessary in the reflector regions.

Calculation of the Macroscopic Group Constant of Pebble Bed Fluoride Salt Cooled High Temperature Reactor Based on the Collision Probability Method

The pebble bed fluoride salt cooled high temperature reactor (PB-FHR) is one of the generation IV nuclear reactors, a lot of study has concentrated on PB-FHR neutronics all over the world. As the most important part in the study work, the macroscopic group constant must be well prepared. The fuel pebble was chosen for the candidate of PB-FHR due to its outstanding, but the double heterogeneous due to its complex structure causes much difficult in the calculation of macroscopic group constant. In this work, an analytical program named Z2D is written to calculate the macroscopic group cross section. In the program, the collision probability method (CPM) was applied to solve the slowing-down integral transport equation, and the macroscopic constant was evaluated with the obtained neutron flux. Also, the recurrence method was introduced to accelerate the computing speed of slowing-down source. The results were compared with those calculated by MCNP, and good agreements were obtained.