Modeling and Design of A New Core-Moderator Assembly and Neutron Beam Ports for the Penn State Breazeale Nuclear Reactor (PSBR)

Open Access
Author:
Ucar, Dundar
Graduate Program:
Nuclear Engineering
Degree:
Doctor of Philosophy
Document Type:
Dissertation
Date of Defense:
March 15, 2013
Committee Members:
  • Kenan Unlu, Dissertation Advisor
  • Kenan Unlu, Committee Chair
  • Kostadin Nikolov Ivanov, Committee Member
  • Maria Nikolova Avramova, Committee Member
  • Thomas Fu Yuan Lin, Committee Member
  • Mahmut Taylan Kandemir, Committee Member
Keywords:
  • PSBR
  • Breazeale
  • TRIGA
  • Neutronic
  • Thermal
  • Hydraulics
  • MCNP
  • ANSYS
  • Fluent
  • Analysis
  • Design
  • Cold
  • Neutron
  • Beam Port
  • CFD
Abstract:
This study is for modeling and designing a new reactor core-moderator assembly and new neutron beam ports that aimed to expand utilization of a new beam hall of the Penn State Breazeale Reactor (PSBR). The PSBR is a part of the Radiation Science and Engineering Facility (RSEC) and is a TRIGA MARK III type research reactor with a movable core placed in a large pool and is capable to produce 1MW output. This reactor is a pool-type reactor with pulsing capability up to 2000 MW for 10-20 msec. There are seven beam ports currently installed to the reactor. The PSBR’s existing core design limits the experimental capability of the facility, as only two of the seven available neutron beam ports are usable. Modeling of the new PSBR design was accomplished by completing neutronic and thermal-hydraulics analyses by using MCNP5, ANSYS Fluent, Gambit, TRIGSIMS, MURE, and Burned Coupled MCNP simulation tool. The finalized design features an optimized result in light of the data obtained from neutronic and thermal-hydraulics analyses as well as geometrical constraints. A new core-moderator assembly was introduced to overcome the limitations of the existing PSBR design, specifically maximizing number of available neutron beam ports and mitigating the hydrogen gamma contamination of the neutron beam channeled in the beam ports. A crescent-shaped moderator is favored in the new PSBR design since it enables simultaneous use of five new neutron beam ports in the facility. Furthermore, the crescent shape sanctions a coupling of the core and moderator, which reduces the hydrogen gamma contamination significantly in the new beam ports. A coupled MURE and MCNP5 code optimization analysis was performed to calculate the optimum design parameters for the new PSBR. The optimum dimensions of the crescent-shaped moderator tank were calculated as 76.2 cm in radius and 48.2 cm in height. All new beam ports were directed to the core center and the optimum distance between the reactor core face and each new beam port was calculated. At optimum distance, which varied between 10 cm and 18 cm for individual beam ports, maximum thermal neutron beam with suppressed fast neutron and gamma components were realized. Thermal-hydraulics analysis of the new design was achieved using ANSYS Fluent CFD code. In the current form, the PSBR is cooled by natural convection of the pool water. The driving force for the natural circulation of the fluid is the heat generation within the fuel rods. The convective heat data was generated at the reactor’s different operating powers by using TRIGSIMS, the fuel management code of the PSBR core. In the CFD modeling, the amount of heat generated by the fuel is assumed to be transferred totally into the coolant. Therefore, the surface heat flux is applied to the fuel cladding outer surface by considering the depleted fuel composition of each individual fuel rod under a reference core loading condition defined as; 53H at 1MW full power. In order to model the entire PSBR reactor, fine mesh discretization was achieved with 22 millions structured and unstructured computational meshes. The conductive heat transfer inside the fuel rods was ignored in order to decrease the computational mesh requirement. Since the PSBR core operates in the subcooled nucleate boiling region, the CFD simulation of new PSBR design was completed utilizing an Eulerian-Eulerian multiphase flow formulation and RPI wall boiling model. The simulation results showed that the new moderator tank geometry results in secondary flow entering into the core due to decrease in the cross-flow area. Notably, the radial flow improves the local heat transfer conditions by providing radial-mixing in the core. Bubble nucleation occurs on the heated fuel rods but bubbles are collapsing in the subcooled fluid. Furthermore, the bulk fluid properties are not affected by the bubble formation. Yet, subcooled boiling enhances the heat transfer on the fuel rods. The CFD predicted bulk fluid temperature in the hot channel of new reactor core was 11 oC higher than the measured bulk fluid temperature in the hot channel of existing reactor core. The maximum bulk fluid temperature in the hot channel is estimated as 88 oC by the CFD model, which is less than the saturation temperature of water coolant at reactor’s operating conditions. In addition, the CFD predicted maximum clad outer surface temperature and calculated maximum fuel temperature in the new PSBR core are estimated as 175 oC and 482 oC, respectively. These temperatures are well below the safety limits of the PSBR. Five neutron beam ports are designed for the new reactor. Four new experimental techniques and facilities are added to the existing Neutron Imaging (NI) and Neutron Transmission facilities in the new beam hall. Triple-Axis Spectrometer (TAS), Conventional and Time-of-Flight (TOF) Neutron Depth Profiling (NDP), Neutron Powder Diffraction (NPD), and Prompt Gamma Activation Analysis (PGAA) are the proposed new techniques. The geometrical configuration, filter and collimator system designs of each neutron beam ports are selected based on the requirements of the experimental facilities. A cold neutron beam port which utilizes cold neutrons from three curved guide tubes is considered. Therefore, there will be seven neutron beams available in the new facility. The neutronic analyses of the new beam port designs were achieved by using MCNP5 code and Burned Coupled Simulation Tool for the PSBR. The MCNP simulation results showed that thermal neutron flux was increased by a factor of minimum 1.23 times and maximum 2.68 times in the new beam port compared to the existing BP4 design. Besides total gamma dose was decreased by a factor of 100 in the new neutron beam facility. Therefore, this study assures both the thermal-hydraulics safety and neutronics performance improvements in the new PSBR design.