Innovative LWR Simulation Tool for the Nuclear Renaissance in the UK

Lead Research Organisation: University of Liverpool
Department Name: Mech, Materials & Aerospace Engineering


The government of the United Kingdom has signed the contracts for the construction of Hinkley Point C, the first nuclear power plant since 20 years as a first step for the Nuclear Renaissance in the UK. In a BBC article, the major points for the construction have been given as Hinkley C will deliver 7% of our electricity when most other nuclear power stations will be closed down. The huge project will provide an economic stimulus.
As a consequence of this decision, the nuclear industry in the UK is facing significant challenges in all engineering disciplines around the new build program as well as the following operation of a fleet of light water reactors (LWRs) instead of the well-known gas cooled reactors of the past. This means the UK has to educate a body of competence for operation, regulation, and consultancy for the construction and the operation of this reactor type. When the reactors are put into operation a part of the reactor core has to be renewed every year while the other part of the core has to be rearranged to make the optimal, economic use of the fuel assemblies. Within this process the operator has to proof each time that the new core configuration is in the envelope of the regulation and the operation will be safe. This proof is mastered by extensive computer simulations. Within these simulations, it has to be proven that the safety limits for each fuel pin in the reactor can be maintained.
At present, the industrial simulation for the fuel management is based on the first generation simulation tools which are now almost 20 years in operation and the evolutionary development of these tools. This scheme is robust, but it lacks in representing the coupling between the different physical phenomena which influence each other. The uncertainty of the results caused by the current limitations has been taken into account by increasing the safety margin to the operational limits which can result in reduced economic performance. The nowadays available computational resources open the possibility for significantly improved solutions. New methods are under development in several other countries. However, many of this high performance 'brute force' solutions are time-consuming and pose requests to the computational environment. This makes these simulation schemes highly interesting from the scientific point of view, but the solutions are not ideal for the industrial application due to the time as well as the computational demand but also due to the high complexity for the user. In contrast to these brute force solutions, I propose a smart approach, which on the one hand can overcome the limitations of the brute force simulation but on the other hand, will deliver a comparable quality of the result in the zones of high interest. The idea is to couple different scientific state-of-the-art tools of the second generation to create a simulation tool which can provide a locally detailed resolution of the reactor fuel structure inside the fuel assembly. This tool will be linked to specific codes for the simulation of the behaviour of the fuel pin inside the fuel assembly. The computational efficiency can be achieved by the focusing the efforts to the zones of interest where the fuel rods are closest to the safety limits.
This approach using different spatial resolutions and coupling different physical simulations for different spatial resolutions has the potential to improve the accuracy of the prediction of the fuel rod like it is required to improve the industrial simulation for nuclear reactors of the nuclear renaissance. It will help to reduce unnecessary high safety margins by improved covering of the underlying physics of the simulations. This will lead to improved economic performance as well as improved safety of the operation and thus reduce the cost of electricity production.

Planned Impact

Simulation of the nuclear reactor's core is the high priority task on all the stages of the nuclear facilities' life-cycle, starting from the design over the operational management to finally the decommissioning. Therefore the accurate modelling of the physical processes in the nuclear power plants is the vital necessity. The quality, as well as performance of the calculations, influence not only on the safety of the nuclear power plant but also on its economic efficiency.
The main beneficiaries of the proposed project are groups in academic as well as in industrial organisations and regulatory bodies performing safety and economical estimations of the nuclear power plants. Another possible community which can gain the benefit from this research project is technical support organisations. Finally, in the case of industrial application of the proposed ideas, the benefit from this research will approach the broad public in the form of lower electricity prices and higher shareholder values of the production companies.
The project is, on the one hand, a way to create new IP for the UK. On the other hand, it supports the UK in being an interesting market for investment and being attractive for students from abroad. Thus attracting bright minds to move to the UK to support the development of an innovative nuclear industry.

Bringing this proposal for a future innovative core simulation into operation will help to develop an advanced tool essential for the specialists in the field of nuclear operation and safety, for nuclear power plants operators and for other specialists in the field of nuclear engineering. They can benefit from this project through using more advanced and reliable computational methods for prediction of the NPP's behaviour. The project will lead to more robust calculation schemes which are more user-friendly than the highly scientific high fidelity methods. Since the advanced calculations are performed for only one or a few assemblies, it is possible to obtain results of the calculations with acceptable effort (in time as well as in computational demand). From the practical point of view, computation should be possible on a desktop computer without the massive use of supercomputers or computer clusters. It is important for industrial application of the codes. Application of the developed program complex on the different stages of the NPP's life cycle will increase its nuclear safety and economic efficiency. Thus, the final consumers of the electricity can benefit from this project through increased safety and economic efficiency of the nuclear power plants which can lead to reduced electricity prices or increased shareholder values of the operating companies. Both will be achieved by reducing unnecessary safety margins which lead to improved economic of the nuclear electricity production.
Description The efficient and safe operation of future nuclear power plants is of high interest for the society. Nowadays, the main approach for the safety evaluation of the nuclear reactors is computer-based modelling & simulation which is able to simulate different phenomena of interest within the reactor core and to evaluate operational as well as safety-related parameters. The main operational limits and safety criteria in the LWR reactor design and operation are applied to the individual pins. However, currently, the industrial codes which are widely used for reactor evaluation are running on the level of the fuel assembly. Therefore, special conservative safety factors are applied to ensure that the key criteria are within safety margins. This approach tends to lead to excessive over-conservatism and prevent more efficient operation of the reactor. One of the solutions can be full core pin-by-pin coupled multiphysics simulation of the whole reactor core. Unfortunately, this approach is very computationally expensive and time-consuming while growing exponentially with the size of the problem. Full core computations using transport solvers, therefore, require, as a rule, a significant computer cluster. Transient simulations are even more expensive and as such can utilize several weeks of cluster time for a simulation of a few seconds of real time of, for example, a reactivity insertion accident. Thus, this approach is neither widely used nor promising for industry application, where the time of simulation is essential. Therefore, another hybrid approach was proposed which allows performing reactor safety calculations using a hybrid multiscale and multi-physics approach without involving massive high-performance simulations while keeping the accuracy on a high level.

In the proposed approach, the whole core calculation is still based on the classical approach of a nodal core simulator (fuel assembly homogenized neutron physics, and a representative fuel rod and a coolant channel), the region of interest (hottest fuel assembly) is resolved down to the pin level in each of the used codes: a transport code with boundary conditions from the full-core calculation for the neutronic solution, a fuel performance code, and a subchannel analysis code. Depending on the quality of the modules (fuel thermodynamics and fluid dynamics) of the core simulator, the typical representative pin and fuel channel can either be modelled in the core simulator or by a multiscale link to the subchannel and fuel performance code. The result will be a computational model of the reactor core, where the hottest fuel assembly (containing the limiting structures) is resolved in all details while the other, less limiting, fuel assemblies are treated as homogenized systems using the already proven models and approximations following the approach to invest the computational power into the area of highest interest. Thus, it provides us with the opportunity to choose where to invest the higher computational effort into higher fidelity and where the current approach is sufficient. The embedding of the hybrid multiscale approach and the additional higher-quality tools (the fuel performance code and the subchannel analysis code) into a regular nodal core simulator will offer the opportunity to apply the higher-fidelity tools when these are required and when it is worth investing the additional computational effort.

As of March 2020, the following results were obtained.
The neutron transport solver which is currently under development by the author was successfully extended for the case of the unstructured, reactor specific meshes allowing to consider both hexagonal and rectangular geometries. The neutron transport solver was successfully tested and verified for the different geometries and data sets. The results of the verification for the single cells was published (see publication section) while the results for the fuel assemblies were submitted to the peer-reviewed journal and are under review at the moment.
Another important direction of the investigation is the coupling of the coarse mesh solver (DYN3D) with the developed neutron transport solver via boundary conditions obtained from the DYN3D. In order to achieve this task, the DYN3D code was modified and the necessary interface for the extraction of the appropriate information was created and introduced into the code. Initial assessment of the proposed technique demonstrates the possibility to reconstruct the powers within fuel assembly with high accuracy. The technique is still being tested at the moment and the findings will be published in the peer-reviewed journal.
Finally, different technologies for multiphysics coupling were considered. HLA (high-level architecture) and MUI (multiscale universal interface) were chosen as most promising. Both of them have their strong and weak sides relating to the project. For example, implementations of the HLA are mostly commercial software with the closed source code while MUI would require updating and changes of the source code before its integration into the project. The final decision on the methodology and software which will be used for the multiphysics coupling will be made after discussion with the specialists from the VEC.

As of March 2021 the following results were obtained:
Different methods were investigated for the pin-power reconstruction within the selected fuel assembly. Obtained results demonstrate that the scheme involving the neutron transport solver significantly outperform the currently used method. Python API was developed for the neutron transport solver allowing for faster and more flexible model generation and smooth data transfer between LOTUS and DYN3D and any other codes which can be potentially coupled in the future. The results of these activities will be published soon in peer-reviewed journals. CTF and FLOCAL thermal-hydraulics validations and verifications within the multiscale and multiphysics software development have been performed to evaluate the accuracy and methodology available to obtain thermal hydraulics at the rod level in both simulation codes. These validations and verifications have proved that CTF is a highly accurate subchannel code for thermal hydraulics. In addition, these verifications have proved that CTF provides a wide range of crossflow and turbulent mixing methods, while FLOCAL in general provides the simplified no-crossflow method as the rest of the methods were only tested during its implementation into DYN3D. Finally, the one-way coupling was performed between DYN3D and CTF which allowed clarifying the structure of the data which should be passed between two codes. Currently, it is implemented using temporary text files, however, the aim is to couple all the codes using in-memory data transfer.

As of March 2022 the following results were obtained:
Coupling of nodal code DYN3D and subchannel code CTF was performed both for single fuel assembly and mini core of the light water reactor. The comparison with the results of the standard DYN3D solver FLOCAL was performed. Coupling with CTF allows for better predictions of important thermal-hydraulic parameters of the system. All necessary interfaces for coupling neutron transport solver and subchannel code were developed and tested. Further work on a coupling of the advanced neutron transport solver LOTUS and nodal code DYN3D led to the development of an innovative method for pin-power reconstruction within fuel assembly of interest. Combined application of the neutron transport solver, subchannel code and nodal code demonstrated the possibility to predict important physical parameters with high accuracy. All findings will be published in peer-reviewed journals in due course.

As of March 2023, the following progress was made on the project:
Nodal code DYN3D, subchannel thermohydraulics code CTF and neutron transport solver LOTUS were coupled in the framework allowing to run of multiphysics simulations in the zones of interest of the reactor core. The results of simulations demonstrated significant differences in the pin powers between DYN3D-CTF coupling and DYN3D-LOTUS-CTF coupling. These findings will be published in a peer-reviewed journal in the next few months.
Exploitation Route These ideas and the vision of the roadmap for future industrial nuclear reactor core simulation in the U.K. to support the nuclear renaissance were published in the open-access journal Energies (see Publications section). The main ideas of the work were incorporated into the BEIS Digital Reactor Design and Advanced Fuels programmes (see Narrative Impact section for more details).
The development of the advanced neutron transport solver LOTUS led to its application for machine learning methods (project "Application of machine learning technologies for the acceleration of massive high fidelity multiphysics simulations in nuclear engineering" funded by NNL).
Further funding was secured in the form of the PhD grant supported by EPSRC and NNL "Application of advanced simulation tools for multi-physics evaluation of the SMR cores". This PhD project is a logical continuation of the Fellowship grant. The outcomes of this Fellowship will be used for multiphysics analysis of the small modular reactors core.
Sectors Digital/Communication/Information Technologies (including Software),Energy

Description The research idea of the Fellowship was directly incorporated into the BEIS Advanced Fuels -Reactor Physics national programme invitation to tender. The published article (see Publications section) on the roadmap for future industrial nuclear reactor core simulation in the U.K. (an essential part of which is formed by the findings of the Fellowship), forms a major part of the roadmap for the future Digital Reactor Design programme. In particular, the requirements given below have been identified in requirement-capturing sessions within the BEIS national programme on Digital Reactor Design, with stakeholders from government, a regulator, and different levels of industrial players taking part. The conclusions were on the following points, which drive the demand for future development: - There is a demand for the development of a system architecture for code coupling to facilitate the transfer of high-fidelity information/data between functions of the whole reactor system; - There is a need for code coupling to extend to the high-fidelity information, potentially down to the atomic scale, for some mission-critical components of the system, particularly within the reactor core, but extending to other areas where such increased fidelity can demonstrate a clear value (either an economic or a safety value); - The code coupling should form part of an overall integrated software framework that is user-friendly and reduces or avoids knowledge requests on specific details from the user
First Year Of Impact 2018
Sector Digital/Communication/Information Technologies (including Software),Energy
Impact Types Policy & public services

Description Application of advanced simulation tools for multiphysics evaluation of the SMR cores
Amount £132,000 (GBP)
Funding ID 2748903 
Organisation Engineering and Physical Sciences Research Council (EPSRC) 
Sector Public
Country United Kingdom
Start 09/2022 
End 09/2026
Description Application of machine learning technologies for the acceleration of massive high fidelity multiphysics simulations in nuclear engineering
Amount £25,000 (GBP)
Funding ID GC_587 
Organisation National Nuclear Laboratory 
Sector Public
Country United Kingdom
Start 12/2021 
End 03/2022
Title Current Coupling Collision Probability Method with Expansion of Flux by Orthogonal Polynomials 
Description The Current Coupling Collision Probability (CCCP) method is a well-known method for the solution of the neutron transport equation. It combines the ideas of the interface current method and collision probability method. Space, where the transport equation should be solved, is subdivided into the set of elements which are further divided into the flat-flux regions. All the space elements are coupled by interface currents and all these elements are internally treated by collision probabilities (CPs). Traditional CCCP methods use flat flux approximation to solve the transport equation. This means that the neutron flux within computational regions is assumed to be constant. Therefore, a very detailed discretization of a region of interest is required to achieve accurate results. This leads to an increase in calculation time since the computation time grows rapidly with additional regions. The developed model is as an extension of the well-known CCCP method. In contrast to the traditional collision probability (CP) method, the flux in the calculation regions is expanded using orthogonal polynomials of two variables up to the 2nd order. It allows reducing the number of calculation regions while maintaining high accuracy. This technique was previously tested for the case of the regular hexagonal lattice. Now, the model was extended for the case of the unstructured mesh. In the case of the regular lattice, the coefficients for the orthogonal polynomials can be calculated analytically, while in the case of the unstructured mesh, the corresponding coefficients should be evaluated numerically. To achieve this, the Gramm-Schmidt procedure combined with a triangulation algorithm is applied. The comparison of this new method to the flat flux approximation demonstrates either an improved quality of the results for identical cell discretization or significantly increased computational efficiency to achieve a comparable accuracy. 
Type Of Material Computer model/algorithm 
Year Produced 2019 
Provided To Others? No  
Impact There is no notable impact from the developed model outside of our research group at the moment. However, within our research group, we realised that using the advanced neutron transport solver with the expansion of flux by orthogonal polynomials reduces the computational time significantly while delivering the results comparable with Monte Carlo simulations. Therefore, this newly developed model will have a high impact on the following progress of the Fellowship. 
Description HZDR Collaboration 
Organisation Helmholtz Association of German Research Centres
Department Helmholtz-Zentrum Dresden-Rossendorf
Country Germany 
Sector Academic/University 
PI Contribution The contribution to this collaboration is informing the partner about the current progress of the Fellowship, key findings and outcomes.
Collaborator Contribution HZDR contributed to the project by providing an academic license of the DYN3D development version including the source code. The DYN3D code is one of the reference codes of the European code platform NURESIM and is widely applied in scientific institutions, regulatory bodies, and industry.
Impact The collaboration in the frame of the current Postdoctoral Fellowship commenced only several months ago. Therefore, there are no significant outcomes available at the moment.
Start Year 2018
Description KM Talk NNL 
Form Of Engagement Activity A talk or presentation
Part Of Official Scheme? No
Geographic Reach National
Primary Audience Industry/Business
Results and Impact In this KM (Knowledge Management) talk organised by NNL, I gave an overview of the current status of the project "Innovative LWR Simulation Tools for Nuclear Renaissance in the UK" financed by EPSRC and supported by NNL. Although the talk was given online, the attendance was pretty good with 43 people joining the session. The talk sparked questions and discussion afterwards and further information on the topic was requested by some of the NNL colleagues.
Year(s) Of Engagement Activity 2021
Description Meeting with NNL 
Form Of Engagement Activity Participation in an activity, workshop or similar
Part Of Official Scheme? No
Geographic Reach National
Primary Audience Industry/Business
Results and Impact A workshop with colleagues from NNL took place on 2nd February 2022. There were about 10 people including NNL specialists in reactor physics, postgraduate students from the University of Liverpool, NNL Senior Reactor Physicist and RAEng Chair in Emerging Technologies. The following topics related to the theme of the Fellowship were discussed:
1. Validation and Verification of Neutron Transport Solver LOTUS.
2. Coupling of Subchannel Analysis Tools with Advanced Multiscale Core Simulations
The ideas and achievements of the Fellowship were discussed with NNL colleagues. Presentations sparked questions and intensive discussion afterwards. Plans for further collaboration were made. NNL demonstrated an interest to invest resources in further research activities on this topic.
Year(s) Of Engagement Activity 2022
Description Meeting with Paul Bryce (EDF Energy) 
Form Of Engagement Activity Participation in an open day or visit at my research institution
Part Of Official Scheme? No
Geographic Reach Local
Primary Audience Other audiences
Results and Impact Paul Bryce - the Lead Technologist, Reactor Physics from Nuclear Technology Branch EDF Energy visited the University of Liverpool. He was informed about the Fellowship's main ideas and activities. The discussion led to the agreement about further meetings and a possible one-day seminar/workshop for the informing the EDF reactor physics team about the progress of the project.
Year(s) Of Engagement Activity 2018
Description Meeting with Paul Bryce (EDF UK) 
Form Of Engagement Activity Participation in an open day or visit at my research institution
Part Of Official Scheme? No
Geographic Reach Local
Primary Audience Industry/Business
Results and Impact In December 2019, a meeting was held at the University of Liverpool with Paul Bryce from EDF UK. At the meeting, issues related to ongoing research were discussed, and a plan was made for the further implementation of the project.
Year(s) Of Engagement Activity 2019
Description Nuclear Academics Meeting 2019 
Form Of Engagement Activity Participation in an activity, workshop or similar
Part Of Official Scheme? No
Geographic Reach International
Primary Audience Other audiences
Results and Impact Nuclear academics and industrial representatives from the UK and other countries (Japan, US, etc.) attended Nuclear Academics Meeting 2019 at Bangor University, Bangor, UK. The discussion of the fellowship with Dr Eugene Shwageraus (Cambridge University) led to the idea to organise workshop/seminar on the fellowship topic at Cambridge University. The discussions with other industry/academics attendees (Tetsushi Hino, Iulia Ipatova) demonstrated an interest in the related subject area.
Year(s) Of Engagement Activity 2019
Description US-UK Modelling & Simulation Best Practice Workshop 
Form Of Engagement Activity Participation in an activity, workshop or similar
Part Of Official Scheme? No
Geographic Reach International
Primary Audience Other audiences
Results and Impact US-UK Modelling & Simulation Best Practice Workshop brought together the specialists in the nuclear reactor modelling and simulation from the US and the UK. The seminar brought together specialists from US and UK national labs, universities and industry. During the workshop, I was involved in a number of discussions with the nuclear specialists both from US and UK, informing them about the Fellowship's research ideas and current progress of the project. In particular, the agreement about further possible collaboration was achieved with Dr Benjamin Collins from ORNL, US.
Year(s) Of Engagement Activity 2019