Nuclear Power Technology Development

IAEA Coordinated Research Projects (CRP)

International co-operative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through Co-ordinated Research Projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes.

Application of Computational Fluid Dynamics (CFD) Codes for Nuclear Power Plant Design

The Coordinated Research Project (CRP) addresses the application of Computational Fluid Dynamics (CFD) computer codes to the process of optimizing the design of water cooled Nuclear Power Plants (NPP). Following a number of initiatives within IAEA where CFD codes are applied to a wide range of situations of interest in nuclear reactor technology, the CRP intends to constitute a systematic framework for the consistent application of those codes. Namely, the CRP will contribute to establish a common vision in relation to the capabilities of CFD codes and their qualification level. The CRP is also expected to provide a roadmap strengthening the application domain of the related technology. Blind analyses (i.e. code calculations performed without having access to experimental data) performed by participants will allow the achievement of the proposed objectives.

Understanding and Prediction of Thermal-Hydraulics Phenomena Relevant to SCWRs

The Super-Critical Water-cooled Reactor (SCWR) is one of the innovative Water Cooled Reactor (WCR) concepts mainly for large scale production of electricity. By utilizing high core outlet coolant temperature, the SCWR is expected to achieve much higher thermal efficiencies than those of current WCRs, and thereby promise improved economics.

The objective of the CRP is to improve the understanding and prediction accuracy of thermal-hydraulics phenomena relevant to SCWRs and to benchmark numerical toolsets for their analyses. Several key phenomena, such as heat transfer, pressure drop and flow stability, have been identified as crucial to the success in developing SCWRs. Experimental and analytical information on these phenomena is being generated at several Member States and can be shared with others to advance the technology.

More information is available from the CRP website

Prediction of Axial and Radial Creep in Pressure Tubes

Pressure tube deformation is a critical aging issue in operating Heavy Water Reactors (HWRs). According to the service year, horizontal pressure tubes have three kinds of deformation: diametral creep leading to the flow bypass and the penalty to critical heat flux for fuel rods, longitudinal creep leading to the interference of feeder pipes and/or with fuelling machine, and sagging leading to the interference with in-core components and potential contact between the pressure tube and calandria tube.

The CRP scope includes the establishment of a database for pressure tube deformation, microstructure characterization of pressure tube materials collected from HWRs currently operating in Member States and development of a prediction model for pressure tube deformation.

More information is available from the CRP website

Heat Transfer Behaviour and Thermo-hydraulics Codes Testing for SCWRs

The higher coolant temperatures proposed for SCWR systems imply fuel cladding temperatures greater than current nuclear reactor operating experience. Because of enhanced heat transfer for supercritical flows and the use of new cladding materials with low corrosion rates, it is necessary to have precise information for establishing both the neutronic and the thermal limits. Consequently, in developing SCWR designs, experimental data for the convective heat transfer from fuel to coolant, covering a range of flow rate, pressure and temperature conditions, are required. The collection, evaluation and assimilation of existing data, as well as conducting new experiments for the attainment of needed data are necessary to establish accurate methods and techniques for the prediction of heat transfer in SCWR cores.

Validated thermo-hydraulic codes are required for the design and safety analyses of SCWR concepts. Existing codes for water-cooled reactors need to be extended in their application and improved to model phenomena such as pressure drop, critical flow, flow instability behaviour and transition from super-critical to two-phase conditions. The appropriate predictive models for computing the heat transfer to super-critical fluids need to be incorporated into the codes, and the codes need to be tested and validated.

More information is available from the CRP website

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

Computer codes used for the analysis of design basis events have been validated against integral and/or separate effects tests, whereas in the case of severe accident computer codes it is rather impossible, or at least quite expensive, to carryout a validation exercise against integrated experiments. Consequently, the code capabilities have to be assessed based on benchmarking against other severe accident computer codes. In view of this, a benchmarking exercise becomes necessary to assess the results from various computer codes to provide an improved understanding of modelling approaches, strengths and limitations. The exercise could also suggest ways to overcome code limitations and thereby increase the confidence in severe accident code predictions. A benchmarking exercise encompassing the various severe accident codes in use within the HWR community is important not only for providing confidence in the overall performance of the codes but also for the reduction of uncertainties in their predictions.

More information is available from the CRP website

Improved Understanding of the Irradiation Creep Behaviour of Nuclear Graphite

South Africa, China, United Kingdom, United States, Japan, South Korea, Ukraine

Isotropic and near-isotropic nuclear grade graphites are used as the nuclear moderator and major structural components of numerous existing power reactors as well as the Gen. IV Very High Temperature Reactors (VHTR), such as the Next Generation Nuclear Plant (NGNP) and the Pebble Bed Modular Reactor (PBMR). During reactor operation graphite core components and core support structures are subjected to complex stresses such as combined loading from neutron irradiation induced dimensional change and thermal gradients. Moreover, static and seismic stresses act on the core components. Stresses in the graphite core are relaxed by irradiation induced creep, and thus it is important to be able to confidently predict the irradiation induced creep strain in a component as a function of dose, temperature, and stress.

Current understanding of graphite irradiation induced creep is based on experimental data from the 1980's and earlier, that only extends to moderate dose and temperature. However, with currently operating reactors reaching high doses, and proposed reactor designs anticipating high neutron doses and temperatures, the need for new creep experiments has become apparent. Furthermore, new graphites have been developed for Gen. IV reactors currently being designed. In response to these needs several member states are planning new irradiation creep experiments to extend the existing database or gather data on the new graphites being used for Gen. IV reactors. Moreover, the development of improved codes and models for the behaviour of existing graphite reactor cores requires the development of improved irradiation creep models.

The CRP brings together scientists and engineers from numerous Member States, all of whom are involved in graphite core assessment, creep experiment designs, or modelling irradiation induced creep in graphite, and will enable comparisons of the relative merits of the various creep models and provide valuable input to the design of future creep experiments.

Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU Reactor Vessel

China, France, India, Japan, Republic of Korea, Russian Federation, and United States of America (two institutes)

The objective of the CRP is to improve the Member States' analytical capabilities in the field of fast reactor in-vessel sodium thermal hydraulics. In particular, the CRP addresses the natural convection behavior of the coolant in the reactor vessel of a sodium cooled fast reactor.

In particular, the CRP participants will perform benchmark exercises focusing, in a first stage, on the numerical simulation of the sodium stratification measurements performed in the MONJU reactor vessel during the original start-up experiments. For the first stage of the CRP, the participants will analyze the sodium thermal stratification effects in the MONJU reactor vessel upper plenum after a plant strip test conducted in December 1995 with the reactor at 45% thermal power level simulating an abnormality in the condenser as triggering event.

The CRP's specific research objectives for this first stage are summarized as follows:

  • review of the detailed description of the boundary conditions of the above mentioned test (e.g. geometrical data, flow rates and temperatures at core outlet, etc), as well as of all the experimental data obtained (various sodium temperature distributions in the upper plenum) and specification of the benchmark models
  • validation of various multi-dimensional fluid dynamics codes in use in Member States through simulation of sodium cooled fast reactor outlet plenum temperature distributions and comparison with experimental data
  • identification of weaknesses in current methodologies (e.g. with regard to turbulence models, reactivity feedback models, etc) and of the R&D needs to resolve the identified open issues

Analyses of and Lessons Learned from the Operational Experience with Fast Reactor Equipment and Systems

The CRP on "Analyses of and Lesson Learned from the Operational Experience with Fast Reactor Equipment and Systems" was aimed at contributing to the preservation of the feedback from the commissioning, operation, and decommissioning of experimental and power sodium cooled fast reactors. The specific objectives of the CRP were to:

  • preserve the feedback from commissioning, operation, and decommissioning experience of experimental and power fast reactors
  • enable easy access to the information from this feedback
  • produce synthesis reports of lesson learned from commissioning, operation, and decommissioning of experimental and power sodium cooled fast reactors

The output of the CRP will be an IAEA Technical Report summarizing the results of the data retrieval and archiving work, as well as the results of the synthesis efforts, of lesson learned and recommendations.

Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems (ADS)

Several countries with nuclear programmes are considering ADS systems as a method to implement nuclear waste transmutation in the scope of their nuclear waste management strategies. The CRP is advancing the Member States' efforts towards designing a demonstration facility by providing the information exchange and collaborative research framework needed to ensure that the tools to perform detailed ADS calculations are available.

The main objective of the CRP is to improve the present understanding of the coupling of ADS spallation sources with multiplicative sub-critical nuclear system. The CRP is addressing all major physics phenomena related to the spallation source and its coupling with the sub-critical system. Integrated calculation schemes are used by the participants to perform computational and experimental benchmark analyses. In a previous IAEA-CRP on "Use of Th-based Fuel Cycle in ADS to Incinerate Pu and to Reduce Long-lived Waste Toxicities" reactor physics benchmark calculations on ADS with fixed external neutron sources have been performed. Comparison of the results of this CRP shows that large discrepancies exist both related to the use of different methods and data. By including comparisons with integral experiments, the current CRP is contributing to the clarification of these discrepancies and validate also those results for which satisfactory agreement was reached in the previous CRP.

Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHENIX End-of-Life Experiments

China, France, India, Japan, Republic of Korea, Russian Federation, Switzerland, and United States of America (two institutes)

The overall objective of the CRP is to improve the Member States' analytical capabilities in the field of fast reactor simulation and design, with particular emphasis on temperature and power distribution calculations, and the analysis of sodium natural circulation phenomena.

The specific research objectives of the CRP are:

  • to perform preparatory analyses for two PHENIX EOL tests
  • to perform blind calculations prior to the tests
  • to perform the post-experiment analyses

The two PHENIX End-of-life tests are "Control Rod Withdrawal Test" and "Sodium Natural Circulation Test". The "Control Rod Withdrawal Test" is performed both in the static and dynamic mode: the comparison of the results allows sensitivity analyses of the two measurement methods, and provides the basis for improving the uncertainty in the determination of power distributions. The objective of the "Sodium Natural Circulation Test" is twofold, including the study of the sodium natural circulation in the primary circuit, as well the determination of the efficiency of natural convection phenomena in the primary circuit, and the qualification of the system codes used to simulate natural convection phenomena.

Benchmark Analyses of an EBR-II Shutdown Heat Removal Test

The CRP addresses Shutdown Heat Removal Tests (SHRT) performed at the Experimental fast Breeder Reactor EBR-II within the framework of the US Integral Fast Reactor development and demonstration programme. The CRP will improve the participants' simulation capabilities in the various fields of research and design of sodium cooled fast reactors through data and codes validation and qualification.

The scope of the CRP is twofold: firstly, validation of the state-of-art liquid metal cooled fast reactor codes and data used in neutronics, thermal hydraulics and safety analyses, and, secondly, training of the next generation of fast reactor analysts through international benchmark exercises.

The Source Term for Radioactivity Release Under Fast Reactor Core Disruptive Accident (CDA) conditions

The CRP will deepen the understanding and perform numerical simulation of the transport mechanism of fission products released subsequent to a CDA in a fast reactor core.

The reference design for the analyses to be performed by the CRP participants is based on the 500 MWe Indian pool-type Prototype Fast Breeder Reactor (PBFR). The CRP participants will identify the fast reactor design parameters which minimize the release of radioactivity into the reactor containment building under the postulated CDA.

SFR: Sodium properties, sodium facility design and safety guidelines

This CRP is intended to address the need of standardization of Na physical and chemical properties, the main rules for designing experimental facilities, good practices and safety guidelines. The CRP will improve the participants' modelling and simulation capabilities in various fields of SFR technology using the same properties, as well as to perform studies on experimental facilities dedicated to various research and development needs. The outputs of this CRP will contribute to an improvement of the future benchmark exercises and of the design of sodium facilities and their safe operation.

New Technologies for Seawater Desalination using Nuclear Energy

Algeria, Egypt, France, India, Indonesia, Kuwait, Rep of Korea, Libya, Morocco, Pakistan, Syria, UK, and USA

This CRP focuses on the introduction of innovative technologies which may help making nuclear desalination more safe and economical. The new technologies are expected to enhance the harvesting of waste heat available in nuclear reactors (i.e. waste heat from the condenser of water cooled reactors, or from the precooler and intercooler of High Temperature Gas Reactors HTGR) and utilize it for seawater desalination. New technologies may involve technologies related to the desalination processes such as Low Temperature-Horizontal Tube Multi Effect Distillation (LT-HT MED), others related to the efficient and maximising heat recovery systems such as heat pipes, or the optimization of coupling configuration between nuclear reactors and desalination systems. Additional dimensions of the CRP are to analyze the economics of cogeneration systems (i.e. for electricity and water production), and improve the IAEA DEEP software.

Please contact NENP Technology Development Section - Contact Point if you have any questions.