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dc.contributor.authorFlaspoehler, Timothy Michaelen_US
dc.date.accessioned2013-01-17T20:47:35Z
dc.date.available2013-01-17T20:47:35Z
dc.date.issued2012-09-27en_US
dc.identifier.urihttp://hdl.handle.net/1853/45743
dc.description.abstractIn the following work, the MAVRIC sequence of the Scale6.1 code package was tested for its efficacy in calculating a wide range of shielding parameters with respect to HTGRs. One of the NGNP designs that has gained large support internationally is the VHTR. The development of the Scale6.1 code package at ORNL has been primarily directed towards supporting the current United States' reactor fleet of LWR technology. Since plans have been made to build a prototype VHTR, it is important to verify that the MAVRIC sequence can adequately meet the simulation needs of a different reactor technology. This was accomplished by creating a detailed model of the VHTR power plant; identifying important, relevant radiation indicators; and implementing methods using MAVRIC to simulate those indicators in the VHTR model. The graphite moderator used in the design shapes a different flux spectrum than water-moderated reactors. The different flux spectrum could lead to new considerations when quantifying shielding characteristics and possibly a different gamma-ray spectrum escaping the core and surrounding components. One key portion of this study was obtaining personnel dose rates in accessible areas within the power plant from both neutron and gamma sources. Additionally, building from professional and regulatory standards a surveillance capsule monitoring program was designed to mimic those used in the nuclear industry. The high temperatures were designed to supply heat for industrial purposes and not just for power production. Since tritium, a heavier radioactive isotope of hydrogen, is produced in the reactor it is important to know the distribution of tritium production and the subsequent diffusion from the core to secondary systems to prevent contamination outside of the nuclear island. Accurately modeling indicators using MAVRIC is the main goal. However, it is almost equally as important for simulations to be carried out in a timely manner. MAVRIC uses the discrete ordinates method to solve the fixed-source transport equation for both neutron and gamma rays on a crude geometric representation of the detailed model. This deterministic forward solution is used to solve an adjoint equation with the adjoint source specified by the user. The adjoint solution is then used to create an importance map that can weight particles in a stochastic Monte Carlo simulation. The goal of using this hybrid methodology is to provide complete accuracy with high precision while decreasing overall simulation times by orders of magnitude. The MAVRIC sequence provides a platform to quickly alter inputs so that vastly different shielding studies can be simulated using one model with minimal effort by the user. Each separate shielding study required unique strategies while looking at different regions in the VHTR plant. MAVRIC proved to be effective for each case.en_US
dc.publisherGeorgia Institute of Technologyen_US
dc.subjectTritiumen_US
dc.subjectReaction rateen_US
dc.subjectNeutral particleen_US
dc.subjectSINBADen_US
dc.subjectHTGRen_US
dc.subjectVENUSen_US
dc.subjectPCAen_US
dc.subjectVariance reductionen_US
dc.subjectFW-CADISen_US
dc.subjectMAVRICen_US
dc.subjectShieldingen_US
dc.subjectDosimetryen_US
dc.subjectSurveillance capsule monitoringen_US
dc.subjectDoseen_US
dc.subjectPhotonen_US
dc.subjectGammaen_US
dc.subjectNeutronen_US
dc.subjectRadiationen_US
dc.subjectTransporten_US
dc.subject.lcshShielding (Radiation)
dc.subject.lcshNuclear engineering Safety measures
dc.subject.lcshRadioactivity Safety measures
dc.titleFW-CADIS variance reduction in MAVRIC shielding analysis of the VHTRen_US
dc.typeThesisen_US
dc.description.degreeMSen_US
dc.contributor.departmentNuclear and Radiological Engineeringen_US
dc.description.advisorCommittee Chair: Petrovic, Bojan; Committee Member: Deo, Chaitanya; Committee Member: Hertel, Nolan; Committee Member: Peplow, Douglasen_US


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