MCNP5 Criticality Safety Modeling and Validation for LWR Fuel Cycle Applications

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dc.contributor.authorHalldin, Niclas
dc.contributor.authorKarin, Rosenqvist
dc.contributor.departmentChalmers tekniska högskola / Institutionen för teknisk fysiksv
dc.contributor.departmentChalmers University of Technology / Department of Applied Physicsen
dc.date.accessioned2019-07-03T13:14:00Z
dc.date.available2019-07-03T13:14:00Z
dc.date.issued2013
dc.description.abstractThis thesis considers criticality safety in the production and storage stages of the light water reactor fuel cycle using the Monte Carlo N-Particle (MCNP) neutron transport code at Westinghouse Electric Sweden (WSE). The fuel production process involves handling aqueous uranium solutions in different types of containers. The criticality safety handbook (KSH) is used as a quick way to assess if the geometry and materials of these containers are critically safe. Recent findings show that the composition of one absorber used in the KSH, polyvinyl chloride (PVC), is not conservative due to additives in the PVC that have not been accounted for. In addition, existing work on the KSH shows that the steel used in the containers cannot always be neglected in the computational models used. In light of this, the KSH is updated with a new PVC composition and steel is included when it is conservative to do so. The effect of the steel thickness on the neutron multiplication factor is evaluated for a number of uranium solutions and geometries. This parametric study also includes other parameters, such as the ratio of different components of the PVC. The results show that the new PVC composition is more conservative than the old one for the cases studied while the addition of steel yields ambiguous results due to the complexity of the different mechanisms involved. In the storage of nuclear fuel in new fuel racks and later in spent fuel pools, criticality must be avoided at all times. This is ensured by criticality safety analyses using e.g. MCNP. In order to be able to rely on the results, the code must be validated for the conditions of the system studied. In this thesis such a validation is done for MCNP5 version 1.51 for criticality calculations for common pressurized water reactor spent fuel storage conditions. This includes systems with soluble boron concentrations up to 3390 ppm and with solid absorbers present. The validation is performed against 101 heterogeneous critical experiments. From the calculations performed for these experiments the accuracy and precision of MCNP5 can be obtained and the upper subcritical limit (USL) may be determined. At WSE a B-term (B = 0.95 – USL) is added to the calculation results in criticality safety analyses to account for the bias of the code used. The validation results show that the smallest and largest B-terms is seen for systems with and without absorbers present respectively. The latter result is unusually large due to the addition of an extra safety margin to account for a lack of data points and may be too conservative.
dc.identifier.urihttps://hdl.handle.net/20.500.12380/180157
dc.language.isoeng
dc.setspec.uppsokPhysicsChemistryMaths
dc.subjectFysik
dc.subjectPhysical Sciences
dc.titleMCNP5 Criticality Safety Modeling and Validation for LWR Fuel Cycle Applications
dc.type.degreeExamensarbete för masterexamensv
dc.type.degreeMaster Thesisen
dc.type.uppsokH
local.programmeApplied physics (MPAPP), MSc
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