Neutronic Design and Analyses of A New Core- Moderator Assembly and Neutron Beam Ports for The Penn State Breazeale Reactor D. Uçar,1 K. Ünlü,1,2 B. J. Heidrich,1 K. N. Ivanov,2 M. N. Avramova,2 Service Provided: Penn State Breazeale Reactor, Neutron Beam Laboratory Sponsors: The Pennsylvania State University Radiation Science and Engineering Center experimental purposes. BP7 is mainly used for neutron Introduction transmission measurements. This port is 12.7 cm The Penn State Breazeale Reactor (PSBR), as a part of below the core center and thus the neutron flux at this Radiation Science and Engineering Center (RSEC), was beam port is significantly lower. Almost all of the other built in 1955 as a research and education hub. It is experimental techniques, i.e. neutron imaging, neutron currently the longest operating research reactor in the depth profiling, detector testing and development etc., United States. The initial reactor design used plate-type are conducted at BP4. However, the high content of the materials testing reactor (MTR) fuel elements with a prompt gamma-rays in these beam ports affects all of 61-cm active fuel length and up to 93% uranium the experiments conducted in the facility. The prompt enrichment. Seven beam ports were built into the gamma-rays are produced by the neutron capture of facility design for analyzing the nuclear properties of the hydrogen in the pool’s water due to the 1H(n,γ)2H materials, determining reactor dynamics, and reaction, which mainly takes place at the sides of the examining the effects of radiation on materials. After D2O tank, see Figure 1. ten years of service, the reactor core design was changed to a TRIGA Mark III. The design conversion to a This study presents a new PSBR core-moderator TRIGA core produced three major advantages for the assembly design and five new beam ports, which would reactor: (1) the reactor power was increased from 200 eliminate all the limitations of the existing design by kW to 1 MW; (2) the reactor used fuel in the low- increasing the number of simultaneously used beam enriched safeguards category since TRIGAs use fuel that ports from two to five and by mitigating the amount of is 20% enriched in uranium, and (3) pulsing capability prompt gamma-rays in the beam port facilities. The was added to the core due to the inherent prompt major constraints of the PSBR are mainly geometric negative feedback characteristics of the TRIGA fuel factors such as available infrastructure in the beam hall, the tower design, geometrical arrangement of the elements, which are a matrix of uranium and ZrH1.6 moderators. Unfortunately, the design conversion also beam ports, and the core and moderator designs. resulted in a partial loss of experimental capability for the facility, such that use of six of the seven beam ports became limited. This is mainly due to the physical differences between MTR and TRIGA fuel element designs. Since the active length of a TRIGA fuel element (38.1 cm) is considerably smaller than the active length of an MTR fuel element (~61 cm), six beam ports, which were aligned with the MTR fuel, are now directed 12.7 or 27.9 cm below the core center. In this existing beam port configuration, only beam port (BP) 4 is located at the core center. In addition, five of the seven existing beam ports could not be properly aligned to the core- moderator assembly after the design change. A schematic drawing of the existing reactor core, D2O tank, graphite reflector, and seven beam ports extended toward the reactor core are given in Figure 1. Therefore, the PSBR is not capable of simultaneously using all the available beam ports with the current configuration of the beam ports and the core- moderator assembly. Only two beam ports, namely BP4 FIGURE 1: A schematic drawing of the PSBR core-moderator and BP7, are coupled with the reactor core for assembly layout with the graphite reflector and the beam ports extended to the reactor core. 1 Radiation Science and Engineering Center, The Pennsylvania State University, University Park, PA 16802 2 Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 Furthermore, the prompt gamma-ray contamination two dry (air-filled) tubes, three fuel-follower control problem and thermal-hydraulics safety of the core are rods (shims, regulator, and safety) and one air-follower design parameters and the neutronic performance of control (transient) rod. The ten graphite rods at the the proposed design are calculated by detailed periphery of the core were removed to achieve proper neutronic simulations and discussed below. coupling with the crescent-shaped moderator tank. The main function of the graphite rods is to enhance Design Considerations neutron economy as well as to decrease the critical The existing core-moderator assembly design is the uranium mass required to achieve criticality. Therefore, main cause of the geometric mismatch of the beam a decrease in the excess reactivity of the reactor after port configuration. The key parameter in the design removal of these rods is expected. On the other hand, process is the calculation of the optimal size and shape the new moderator tank is bigger in size and covers of the moderator tank. The limitations of the PSBR more than half of the core periphery, which will result were previously studied by Butler, who specifically in a positive impact on the excess reactivity of the analyzed the utilization of three moderator tank shapes system. These two competing design changes on the for the PSBR (crescent, horseshoe and rectangular) and excess reactivity of the reactor were analyzed. the geometric arrangement of five new neutron beam ports in a moderator tank [1]. In this study, a crescent- shaped moderator tank is favored since it allows for the simultaneous use of five beam ports. After the selection of the moderator tank shape, the second design step is the proper coupling of the moderator tank with the reactor core in order to eliminate the prompt gamma-ray contamination problem by minimizing pool water at the interface of the core- moderator assembly. This was achieved by keeping the faces of the top and bottom grid plates and the crescent-shaped moderator tank as close as possible (0.62 cm between the core and the moderator tank). The final step in the design process is how to support a new core design with a new reactor tower. The existing reactor core is supported by a tower through the bottom grid plate. The top grid plate is connected to the bottom grid plate. In the new design, the top and FIGURE 2: 3D CAD drawing of the new core-moderator bottom grid plates are equal in size and smaller than assembly and tower design. the existing grid plates. As a result, the tower design will be changed by installing four new support bars and The second part of the neutronic analysis is the two supports plates on top of the core. Figure 2 shows calculation of the optimal configuration of the core- the core-moderator assembly and tower design for the moderator assembly and neutron beam ports, which PSBR after the design changes. will provide high-flux thermal neutron beams with For the redesigned reactor, four thermal and one cold minimal background radiation to the beam port neutron beam ports are proposed for various neutron facilities. The simulation was performed using Monte techniques. Since the cold neutron beam port will Carlo N-Particle ver. 5 (MCNP5) code with sensitivity channel three curved neutron guide tubes, seven studies [2]. First, the design parameters were instruments can be simultaneously used in the beam evaluated. The simulation approach was to make a hall. Four new techniques are planned for the facility: sensitivity study by changing a parameter of interest, Triple-Axis Spectrometry, Prompt Gamma Activation such as moderator size, while keeping all the other Analysis (PGAA), Convectional and Time-of-Flight (TOF)- parameters unchanged. The effect of the design Neutron Depth Profiling (NDP), and Neutron Powder parameter on the neutronic performance of the new Diffraction (NPD). In addition, the existing neutron beam ports was examined in successive MCNP transmission and neutron imaging facilities will remain calculations. The optimal value of the design parameter available. of interest was selected to yield the maximum thermal to fast neutron flux ratio and minimum gamma-ray Optimization and Performance Analyses of New dose at the exit of the new beam ports. Core-Moderator Assembly Using Simulations After the optimal design parameter values were Design and neutronic simulations of the new reactor calculated, the neutronic performance of the new core were performed with a reference core model reactor with five new beam port designs was analyzed selected as loading 53H, which went critical in May by comparing the neutron and gamma-ray flux 2009. There are 102 fuel elements, ten graphite rods, distributions to the measured and simulated spectra in the existing BP4 and BP7. The filter and collimator systems in each new beam port were selected to removal of the graphite rods, the existing D2O tank and accommodate the requirements of the neutron beam the graphite block, it is expected that the new core- technique for the beam port of interest. moderator assembly design will provide $0.15 higher excess reactivity to the reactor. Reactivity Analysis Evaluation of Optimal Design Parameters The excess reactivity of the new reactor core was analyzed by using MCNP5 and TRIGSIMS (TRIGA The goal for the project is the determination of the Simulator-S), the fuel management code system of the optimal dimensions of the moderator tank, beam ports, PSBR based on Monte Carlo methods [3].