Review of the C-nat(n,gamma) cross section and criticality calculations of the graphite moderated reactor BR1

A review of the experimental data for natC(n,c) and 12C(n,c) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK-CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data

Analysis of 238Pu and 56Fe Evaluated Data for Use in MYRRHA

A sensitivity analysis on the multiplication factor, keffkeff, to the cross section data has been carried out for the MYRRHA critical configuration in order to show the most relevant reactions. With these results, a further analysis on the 238Pu and 56Fe cross sections has been performed, comparing the evaluations provided in the JEFF-3.1.2 and ENDF/B-VII.1 libraries for these nuclides. Then, the effect in MYRRHA of the differences between evaluations are analysed, presenting the source of the differences. With these results, recommendations for the 56Fe and 238Pu evaluations are suggested. These calculations have been performed with SCALE6.1 and MCNPX-2.7e

Blind Benchmark Exercise for Spent Nuclear Fuel Decay Heat

The decay heat rate of five spent nuclear fuel assemblies of the pressurized water reactor type were measured by calorimetry at the interim storage for spent nuclear fuel in Sweden. Calculations of the decay heat rate of the five assemblies were performed by 20 organizations using different codes and nuclear data libraries resulting in 31 results for each assembly, spanning most of the current state-of-the-art practice. The calculations were based on a selected subset of information, such as reactor operating history and fuel assembly properties. The relative difference between the measured and average calculated decay heat rate ranged from 0.6% to 3.3% for the five assemblies. The standard deviation of these relative differences ranged from 1.9% to 2.4%

Neutronic design of MYRRHA reactor hall shielding

The lateral shielding of a 600 MeV proton linear accelerator beam line in the MYRRHA reactor hall has been assessed using neutronic calculations by the MCNPX code complemented with analytical predictions. Continuous beam losses were considered to define the required shielding thickness that meets the requirements for the dose rate limits. Required shielding thicknesses were investigated from the viewpoint of accidental full beam loss as well as beam loss on collimator. The results confirm that the required shielding thicknesses are highly sensitive to the spatial shape of the beam and strongly divergent beam losses. Therefore shielding barrier should be designed according to the more conservative assumptions

Impact of intermediate and high energy nuclear data on the neutronic safety parameters of MYRRHA accelerator driven system

Perturbation of external neutron source can cause significant local power changes transformed into undesired safety-related events in an accelerator driven system. Therefore for the accurate design of MYRRHA sub-critical core it is important to evaluate the uncertainty of power responses caused by the uncertainties in nuclear reaction models describing the particle transport from primary proton energy down to the evaluated nuclear data table range. The calculations with a set of models resulted in quite low uncertainty on the local power caused by significant perturbation of primary neutron yield from proton interactions with lead and bismuth isotopes. The considered accidental event of prescribed proton beam shape loss causes drastic increase in local power but does not practically change the total core thermal power making this effect difficult to detect. In the same time the results demonstrate a correlation between perturbed local power responses in normal operation and misaligned beam conditions indicating that generation of covariance data for proton and neutron induced neutron multiplicities for lead and bismuth isotopes is needed to obtain reliable uncertainties for local power responses

Impact of intermediate and high energy nuclear data on the neutronic safety parameters of MYRRHA accelerator driven system

Perturbation of external neutron source can cause significant local power changes transformed into undesired safety-related events in an accelerator driven system. Therefore for the accurate design of MYRRHA sub-critical core it is important to evaluate the uncertainty of power responses caused by the uncertainties in nuclear reaction models describing the particle transport from primary proton energy down to the evaluated nuclear data table range. The calculations with a set of models resulted in quite low uncertainty on the local power caused by significant perturbation of primary neutron yield from proton interactions with lead and bismuth isotopes. The considered accidental event of prescribed proton beam shape loss causes drastic increase in local power but does not practically change the total core thermal power making this effect difficult to detect. In the same time the results demonstrate a correlation between perturbed local power responses in normal operation and misaligned beam conditions indicating that generation of covariance data for proton and neutron induced neutron multiplicities for lead and bismuth isotopes is needed to obtain reliable uncertainties for local power responses

Neutronic design of MYRRHA reactor hall shielding

The lateral shielding of a 600 MeV proton linear accelerator beam line in the MYRRHA reactor hall has been assessed using neutronic calculations by the MCNPX code complemented with analytical predictions. Continuous beam losses were considered to define the required shielding thickness that meets the requirements for the dose rate limits. Required shielding thicknesses were investigated from the viewpoint of accidental full beam loss as well as beam loss on collimator. The results confirm that the required shielding thicknesses are highly sensitive to the spatial shape of the beam and strongly divergent beam losses. Therefore shielding barrier should be designed according to the more conservative assumptions

Depletion uncertainty analysis to the MYRRHA fuel assembly model

In this work, the objective is to perform an uncertainty analysis on a MYRRHA -Rev.1.6 irradiation cycle study, being applied to a depletion scenario of a single fresh fuel assembly while assuming reflective boundary conditions. Such analysis is statistically based on the application of Wilk’s method of building tolerance limits after 100 depletion calculations were performed with the SERPENT2 code. Due to the computational burden of such type of simulations, this propagation of nuclear data covariances study (allowed by the fast computational performance of SERPENT2) was done at constant power, constant flux and, in a final exercise, at constant power with the addition of fission yield uncertainties (all of these cases employed ENDF/B-VII.1 data). It was observed that while depleting at constant power, the statistical variation of key fission products such as 148Nd is almost not present because of the normalization factor applied to the flux. In contrast, the irradiation at constant flux reveals dependence on burnup. Finally, the added fission yield uncertainties make clear the fact that they directly impact the degree of final uncertainty computed for fission products exemplified by 148Nd and 135Xe important for burnup estimation and reactor operation, respectively