Investigation of Coupled Code Pressurized Water Reactor Simulations Using CTF with Soluble Boron Tracking
Open Access
- Author:
- Biery, Mark Byron
- Graduate Program:
- Nuclear Engineering
- Degree:
- Master of Science
- Document Type:
- Master Thesis
- Date of Defense:
- None
- Committee Members:
- Maria Nikolova Avramova, Thesis Advisor/Co-Advisor
Kostadin Nikolov Ivanov, Thesis Advisor/Co-Advisor - Keywords:
- Reactor Core
Coupled Code
NEM
CTF
Boron
Boron Tracking
Mixed Oxide
MOX
PWR
Dillution
Transient
Benchmark - Abstract:
- For long term reactivity control over a nuclear reactor core fuel cycle, pressurized water reactors make use of chemical shim in the form of soluble boron added to the coolant water. While soluble boron allows for even reactivity control and more uniform fuel burn-up, maintaining uniform distribution of the boron is important to prevent localized transients. Transients that are caused by a local disturbance in the concentration of boron are classified as boron dilution transients. While many studies have been performed to study these types of transients, the choice of existing codes available to simulate soluble boron transport have required tradeoffs to be made. Popularly used system codes such as RELAP5-3D can only simulate one-dimensional boron transport with comparatively simple physical models which neglect important physical characteristics of boron transport in the fluid such as mixing due to cross flow between channels and turbulence effects. On the other extreme, Computational Fluid Dynamics (CFD) codes are capable of modeling boron transport with very high fidelity, but most CFD codes still require a large amount of computational resources to simulate a realistic physical model. Recent work by members of The Pennsylvania State University, Department of Mechanical and Nuclear Engineering, Reactor Dynamics and Fuel Management Group has helped to fill this capability gap. The result is an improvement to the Penn State version of COBRA-TF, PSU CTF by employing a newly developed boron tracking model. The resulting version of CTF is known as CTF-BTM. The implemented boron tracking model uses a Modified Godunov method to solve the boron transport field equation. Although the CTF boron tracking model was rigorously tested at the time it was developed, it has not yet been used in coupled thermal hydraulics and neutronics simulations, which is the aim of this study. The objective of this study is to continue the validation and qualification of the boron tracking model used in CTF-BTM. This is accomplished by first coupling CTF-BTM to the nodal diffusion-based neutronics code NEM. Part II and Part III of the 2007 OECD / NEA MOX / UO2 Core Transient Benchmark are then used to validate the coupled code at Hot-Full Power (HFP) conditions and Hot-Zero Power (HZP) conditions. Close agreement to the benchmark solutions is achieved in both cases. Between the boron tracking and non-boron tracking code versions, the results at HZP were nearly identical with a 0.01 ppm difference of predicted critical boron and a 0.33% deviation from the benchmark reference solution. At HFP conditions, the differences introduced by the boron tracking model were more obvious. The boron tracking model result produced a larger deviation from the benchmark reference solution with a 0.61% deviation versus a 0.02% deviation predicted by the non-boron tracking code version. It was concluded that this larger deviation from the benchmark reference solution was due to the fact that the boron tracking model provides a more realistic treatment of the boron transport behavior by allowing the boron concentration to change with moderator density changes and moderator void formation. Although a soluble boron transient benchmark was not available for this particular core model, a systematic approach is used to build upon the successful steady-state benchmarking of the coupled code. For transient verification, a series of steady-state calculations are first carried out to find the effective multiplication factor of the core at varying levels of coolant boron concentration. Using these data to predict the core reactivity change for a prescribed change in inlet boron concentration, a series of inlet boron concentration transient simulations are carried out. The core response to the boron transient is then compared to the predicted reactivity change. Of five cases simulated, it was found that smaller and more rapid changes in reactivity produced a result that more closely matched predicted reactivity. More gradual and larger changes in reactivity caused fuel heating or cooling to occur which introduced Doppler feedback effects. Although it proved difficult to produce transient simulation results that exactly matched predicted reactivity changes, deviations from predicted reactivity changes could be explained by changes in fuel temperature corresponding to changes in reactivity introduced by the change in coolant boron concentration. This study then culminates in the execution of a postulated post-Small Break Loss Of Coolant Accident (SBLOCA) boron dilution accident (which is considered to be a reactivity insertion accident) and an accompanying sensitivity study. While the scenario is highly idealized in terms of initiating events and assumptions, it provides an example of one of the expected future applications of coupled CTF and three-dimensional neutronics codes in LWR simulations with the introduction of boron tracking capability. In this culminating scenario, a series of simulations are executed where deborated condensate water slugs are inserted into the core. The slugs are formed in the steam generators by condensation following loss of coolant inventory sufficient to maintain natural circulation. In the natural circulation cases, a 1.16 m3 slug is assumed. Following formation of the condensate mass, natural circulation is assumed to resume, carrying the condensate slug into the core. The location of slug entry is varied from the core periphery to the core center along an expected path of travel in the lower plenum. It is found that, of the locations where the slug was expected to enter, the most limiting location was fuel assembly E3 (sub-channel 158). The limiting case was determined by the peak fuel enthalpy in the hottest fuel node, which in this case peaked at 76.61 cal/g. This is well below the enthalpy level expected to cause fuel damage which, for a reactivity insertion accident such as this, is expected to occur at 230 cal/g. After finding the limiting location of condensate slug entry in the natural circulation cases, the case is simulated again with the condensate slug entering the same location with the same volume, but at forced circulation conditions assuming actuation of the corresponding reactor coolant pump. As expected, the core response was more violent than exhibited in the natural circulation cases with a larger power excursion and peak fuel enthalpy of 543.99 cal/g. At this fuel enthalpy level, extensive fuel damage is expected. In the final simulation case, the forced circulation case is executed again, but with a 4 m3 condensate slug volume. The cross sectional area of the slug remains unchanged which gives the slug a longer length. In this case, the heating of the fuel occurred for a longer period of time due to the longer power excursion event. The peak fuel enthalpy in this case was 703.81 cal/g with extensive fuel damage expected. Given the assumptions made in the boron dilution accident scenario simulation cases, the formulation used represents worst case boron dilution accident conditions due to the amount of approximations and simplifications used and the level of compensatory conservatism in the formulation. While many studies have been performed elsewhere on boron dilution accidents for currently operating Pressurized Water Reactors (PWRs) this study is one of few focused on boron dilution simulations in a PWR with a MOX/UO2 core.