Chad L. Pope

By Chad Pope and Edward Lum

The energy released from the nuclear fission process drives thermal expansion and mechanical interactions in nuclear reactors. These phenomena cause changes in the neutron chain reaction which results in further changes in thermal expansion and mechanical interactions. Coupling finite element analysis with Monte Carlo neutron transport analysis provides a pathway to simulate the thermal expansion and mechanical interaction to determine a fundamental parameter, namely, thermal expansion temperature coefficient of reactivity. Knowing the coefficient value allows predictions of how a reactor will behave under transient conditions. Using the coupling of finite element analysis and Monte Carlo neutron transport analysis, the thermal expansion temperature coefficient of reactivity was determined for the Godiva-IV reactor (−2E−05 Δk/k/°C) and the Experimental Breeder Reactor-II (EBR-II) (−1.4E−03 $/°C). The Godiva-IV result is within 3% of the measured result. The thermal expansion and mechanical interactions within EBR-II are sufficiently complex that experimentally measuring the isolated coefficient of reactivity was not possible. However, the calculated result fits well with the integral EBR-II reactivity coefficient measurements. Coupling finite element analysis with Monte Carlo neutron transport analysis provides a powerful technique that gives reactor operators and designers greater confidence in reactor operating characteristics and safety margins.

Part of the book: Finite Element Methods and Their Applications

By Alison Wells and Chad L. Pope

Traditional component pass/fail design analysis and testing protocol drives excessively conservative operating limits and setpoints as well as unnecessarily large margins of safety. Component performance testing coupled with failure probability model development can support selection of more flexible operating limits and setpoints as well as softening defense-in-depth elements. This chapter discuses the process of Bayesian regression fragility model development using Markov Chain Monte Carlo methods and model checking protocol using three types of Bayesian p-values. The chapter also discusses application of the model development and testing techniques through component flooding performance experiments associated with industrial steel doors being subjected to a rising water scenario. These component tests yield the necessary data for fragility model development while providing insight into development of testing protocol that will yield meaningful data for fragility model development. Finally, the chapter discusses development and selection of a fragility model for industrial steel door performance when subjected to a water-rising scenario.

Part of the book: The Monte Carlo Methods

By Chad L. Pope and William Phoenix

A category of reactors called university research and teaching reactors, includes relatively high-power pool-type and low-power solid-core reactors. Many high-power university reactors are largely used for irradiations and isotope production. Their almost constant operation tends to impede student access. A university reactor can be particularly relevant to the university’s mission of preparing well-rounded students who have theoretical knowledge, reinforced by focused laboratory reactor experience. The solid-core Idaho State University Aerojet General Nucleonics (AGN) model 201 reactor operates at such a low power (5 W maximum) that it is not useful for isotope production activities. However, the AGN-201 reactor is well suited for teaching and research activities. The solid-core AGN-201 reactor requires no active cooling system, uses a simple shielding arrangement, and the very low operating power results in trivial burnup providing an operating lifetime exceeding many decades. It is thus worthwhile to examine the Idaho State University AGN-201 nuclear reactor more closely because it offers a wide range of research and teaching capabilities while being widely available to students.

Part of the book: Nuclear Reactors

By Chad L. Pope, Ryan Stewart and Edward Lum

The Experimental Breeder Reactor II (EBR-II) operated from 1964 to 1994. EBR-II was a sodium-cooled fast reactor operating at 69 MWth producing 19 MWe. Rather than using a loop approach for the coolant, EBR-II used a pool arrangement where the reactor core, primary coolant piping, and primary reactor coolant pumps were contained within the pool of sodium. Also contained within the pool was a heat exchanger where primary coolant, which is radioactive, transferred heat to secondary, nonradioactive, sodium. The nuclear power plant included a sodium boiler building where heat from the secondary sodium generated superheated steam, which was delivered to a turbine/generator for electricity production. EBR-II fuel was metallic uranium alloyed with various metals providing significant performance and safety enhancements over oxide fuel. The most significant EBR-II experiments occurred in April 1986. Relying on inherent physical properties of the reactor, two experiments were performed subjecting the reactor to loss of primary coolant flow without reactor SCRAM and loss of the secondary system heat removal without reactor SCRAM. In both experiments, the reactor experienced no damage. This chapter provides a description of the most important design features of EBR-II along with a summary of the landmark reactor safety experiments.

Part of the book: Nuclear Reactors