Key BWR Research Needs
At the very beginning of the BWR Research Hub and Network a workshop was held in Bangor to discuss the research needs surrounding boiling water reactor technology. This page exists to summarise the results of this meeting and will be updated to reflect the changing priorities for BWR research as they develop.
During the Research Needs of Boiling Water Reactors Conference held in October 2016, engineers and scientists from Hitachi-GE gave an overview of what they saw the research priorities should be, given their operational experience with BWRs in Japan. Their presentations are linked here:
- Overview of Research Needs of the BWR (K. Moriya)
- Research Needs for Core and Fuel Design (T. Hino)
- Research Needs for Higher Performance Core and Fuel (H. Soneda)
- Research Needs for Thermal Hydraulics (K. Nishida)
Breakout sessions were held on the second day of the workshop during which attendees discussed how BWR research might complement existing UK nuclear research programmes. They also considered research areas that would be of mutual benefit to the UK and Japan. The results of these sessions are now summarised.
The breakout session comprised discussing and then compiling a list of potential research areas, followed by an attempt to identify the areas of greatest interest to both Hitachi and UK industry and academia. The list was compiled by Mike Thomas (NNL, Reactors Team) who acted as session chairman.
List of potential research areas of mutual benefit
Each item was placed into one or more of the following categories:
- M: Methods/computer code development, including core design, fuel performance and neutronics/thermal-hydraulic coupling.
- C: Fuel cladding research.
- E: Environmental research, including crud investigations, coolant activation , material degradation and environmentally-assisted cracking.
The list of possible research areas now follows:
- Fuel performance analysis methods for higher burnup fuel (e.g., extending the validation database to higher burnups) (M)
- Development and validation of modelling techniques to be used to represent the harder neutron spectrum (cf. a thermal reactor) when operating a Reduced moderation Boiling Water Reactor (RBWR) (M)
- Collaboration on advance numerical modelling methods, such as reduced order modelling, predictive modelling and uncertainty propagation (M)
- Development and use of UK neutronic and T&H BWR methods (e.g., WIMS/PANTHER could be coupled with a T&H/system code such as RELAP) enabling comparisons with Hitachi calculations (M)
- Corrosion of Zr alloy claddings – mechanistic understanding of oxidation under single and two-phase environments, prediction of oxidation rates, development of more resistant alloys (C)
- Post Irradiation Examination (PIE) of active samples, e.g., cladding (C)
Novel clad coatings, e.g., graphene and surface treatments (C)
- Possible internship(s) at Hitachi, focussing on core/fuel modelling of existing technology using Hitachi techniques and methodology (with a view to using these techniques back in the UK using UK methods) (M)
- Further development of RBWR analyses, adding to work already carried out in the UK (M)
- Investigation into failed fuel rod identification techniques using robotics and laser ultrasonic testing (C)
- Comparison between UK (ENIGMA) and Hitachi (PRIME) fuel performance codes, including importing of BWR specific models and development of ENIGMA for BWR operating conditions (already licensed in the UK for Advanced Gas Reactor and PWR purposes) (M)
Atomistic modelling of fuel, e.g., fission gas release, Pellet Clad Interaction, rim effects (M)+ (C)
Chemical release rates of stellite alternatives (E)
- Research into radiation damage of steels (E)
Research into stress-corrosion cracking (including irradiation-assisted SCC of austenitic stainless steel components, SCC during start-up and cool-down, effectiveness of different water chemistries, hydrogen embrittlement) (E)
- Further investigation into BWR crud deposition in appropriate water chemistries and flow conditions (UK work carried out currently on behalf of EPRI) (E)
- Research into corrosion fatigue including applicability of design curves (E)
- Further workshop on development of reactor core design and safety analysis methods and techniques (M)
List of Initial Priorities
- Core design and core analysis methods development, e.g., modelling MOX in ABWR and possibly RBWR (M)
- Possible coupling of balance-of-plant code (e.g., RELAP) with core neutronics code (e.g., PANTHER) perhaps incorporating novel ‘reduced order’ numerical methods (M)
- Possible import of further BWR models into UK’s ENIGMA fuel performance code, followed by UK(ENIGMA) vs. Hitachi(PRIME: Developed and owned by Global Nuclear Fuel)) code comparison (M)
- UK groups currently working on cladding materials → would be beneficial to map this work onto BWR conditions, including microstructural investigations of BWR alloys irradiated in a reactor (as opposed to ion beam irradiation)
This memo summarises the proceedings of the “BWR Thermal Hydraulics” session and was compiled by Richard Stainsby (NNL) who acted as session co-chair.
List of potential research areas of mutual benefit
Whole-Systems Code Developments (from presentation on research needs for BWR high performance core and fuel)
Systems codes are used to analyse the transient performance of reactor systems, including significant elements of the power conversion system, to analyse and demonstrate the behaviour of the whole system during normal operational transients, anticipated operation occurrences (AOOs) and during fault conditions. System codes are used to build up very complicated flow, heat transfer and control logic networks to mimic the plant using fairly simple building blocks. The basic hydraulic building block is essentially a pipe element. Complex structures such as a reactor core are formed by assembling a bundle of such pipe-elements, with transverse cross-junctions if discrete cross flows between the channels are possible. This approximation is reasonable for a set of shrouded fuel elements, as in the ABWR, but it becomes less applicable if taken down to the level of individual subchannels with any given bundle. Further, the move to half-height pins encourages transverse redistribution of flow within bundles.
To date Hitachi have based their analysis of ABWR transients on the TRACG systems code, sometimes coupled with a neutronics model for transients in which resolution of thermal-neutronic feedbacks is important. TRACG is a customised version of the Los Alamos code TRAC, with the customisation performed by General Electric, to produce a version specific to BWRs.
As in the case of standard TRAC, TRACG is based on a two-fluid model of two-phase flow.
The session attendees identified two potential areas for development of modelling for the reactor core:
First, the parallel pipe approximation of the reactor core that is achievable with TRACG misses a lot of detail that occurs at the sub-channel level, particularly in the vicinity of grids and other discontinuities (or continuous features, such as an imbalance in sub-channel powers) which promote transverse flow.
Second, whilst the two-fluid model performs well in regions of low (or no) heat flux, such as in the external piping circuit, this does not accurately capture the behaviour and interaction between the vapour, film and droplet fields within the core sub-channels.
The proposal is to develop a new “core component” that can be linked together with a TRACG model of the rest of the BWR system to resolve both issues. Such a component would be based on a subchannel scale representation of a fuel bundle, with the capacity for the model to include an arbitrary number of parallel bundles. Ideally this core component would be based on a multi-fluid model (with at least three fluids for vapour) of the two-phase flow within the core. The precedent for this type of model already exists in the form of COBRA/TRAC, in which COBRA is a detailed continuum and/or sub-channel-scale code based on a two-fluid, three-field model of two phase flow.
Development of TRACG to replace the two fluid model with a multi-fluid model was considered to be too onerous and would undermine the existing validation basis of the code. If the application of the code is limited to non-core components, the following areas for development were identified:
- A consideration of / or improvement to the treatment of interfacial friction.
- Friction and heat transfer correlations in mixed and natural convection regimes.
Thermal Hydraulics Developments
Test Rigs and instrumentation
“Big” test rig to examine two phase flow behaviour in core channels at representative heat flux, temperature and pressure.
Instrumentation to measure, and ideally visualise:
- Film thickness, interface structure and liquid flowrate measurement (film characterisation) in isothermal, heating and cooling conditions.
- Bubbles and droplets, spatial and size distributions and velocities.
Understanding the structure and behaviour of churn flow
- Turbulence – understanding the relationship between vapour-phase turbulence intensity and droplets for annular dispersed flows.
To facilitate some of the above, it would be good to investigate use of radioactive tracers, or temporarily induced radioactivity in the liquid phase (17O)from a pulsed neutron source to carry out gamma tomography (PET in the latter case) imaging of the distribution of liquid in the two-phase region in real time.
“Small” test rig to investigate same phenomena by optical means in glass.
There is a strong cross over between the needs of the core component model for transient analysis discussed above with the needs of a detailed thermal hydraulics model of the core to carry out coupled neutronics / thermal hydraulics analyses. Such a model will have an improved capability to predict void distribution and power levels/shapes at which critical heat flux occurs.
The best of the generally available codes appears to be based on two-fluid, three-field models. The use of true multi-fluid models appears to be restricted to “academic” codes, so there is scope to implement a true three fluid two-phase model in a core thermal hydraulics code, which could be transferable to other industrial problems where accurate resolution of phase change is important.