Indo-UK Civil Nuclear Collaboration Reactor Design and Development Group

R&D Activities in RDDG for “Plant Design” Reactor physics Fuel design Thermal hydraulics Life extension of coolant channel Core safety Containment thermal hydraulics and safety Reactor structures Vibration diagnostics Instrumentation and control PSA Refuelling technology

Collaborative Projects on Thermal hydraulics Collaboration was based on the availability of experimental facilities at BARC and numerical and modeling abilities at Imperial College. Projects undertaken Phase I: Verification and Validation of CHF prediction and CFD Phase II: Thermal hydraulics of boiling and passive system Phase III: (a) Premature Oscillation Induced CHF (OICHF) (b) New reactor designs from a safety systems perspective, passivity and grace time Collaborating Institutes India Reactor Engineering Division, BARC UK Imperial College London University of Sheffield University of Leeds

Phase-1 Objectives: 1) Measurement of Film thickness and flow rate in Annular flow regime for dryout modelling 2) CFD validation Facilities utilised: (1) Air-water facility (2) High pressure and High temperature facility (CHIL) (3) Counter Current Flow Facility Annular flow Churn flow Liquid film flow rate in steam-water conditions (BARC) Air-Water Facility (BARC) Outcome Sensors and devices for the film flow rate and film thickness developed jointly Liquid film thickness and film flow rate were measured for dryout modelling Data generated in BARC and compared with available models in lirerature

Counter current flow set-up (BARC) Phase-1 (contd..) CFD Validation: Counter current natural circulation - experiments and modelling Set-up created for the measurement of counter current flow phenomenon (BARC) Flow visualisation and flow field studies using PIV system and CFD studies (BARC) Simulation using CFD code (UK) Counter current flow set-up (BARC) DNS simulation (UK) Velocity field measured using PIV (BARC)

Microscopic boiling measurements Phase-II Objective : Dryout modeling (extension of phase –I) Microscopic boiling studies Supercritical water studies Outcome Inter-code comparisons BARC Code, SCADOP developed for analysis of dryout in rod-bundles. UK code, GRAMP Westinghouse code, MEFISTO-T Excellent agreement between the codes and data Measurements on Microscopic boiling carried out and data is being used for validation of the UK Code Bubble growth and departure diameter and velocity were measured. Inter-code comparisons Microscopic boiling measurements Bubble Departure Diameter

SC water facility (University of Sheffield ) Phase-II (contd..) In supercritical water cooled reactors (SCWRs), the fluid does not change phase, but its properties still vary drastically and the density could change up to 7 times in the core for example. Such changes, especially the buoyancy due to density changes may cause significant heat transfer deterioration. Understanding/predicting such phenomena are important for SWCR designs. g x y z 2δ (160) 4δ(160) 16δ(512) Tc Th Flow Direction SC water facility (BARC) SC water facility (University of Sheffield ) prediction with BARC data Outcome In a vertical channel, flow structures vary significantly from the hot to the cold surfaces. Turbulence (hence heat transfer) is heavily modified in the various flow cases in comparison with ‘normal ‘ flow (black)

Phase-II (Contd...) Improvement of the boiling model and CHF prediction (Imperial College and Leeds) Joint work between Imperial College and University of Leeds Advances in multiphase turbulence and population balance combined with more mechanistic wall boiling model Validation with sub-cooled boiling flows underway Validation against CHF data provided by BARC underway Boiling flow prediction CHF data CHF data (BARC) for validation Boiling void prediction (UK)

Phase-III (a) Premature-Oscillating Induced CHF (OICHF) Experiments are in progress to quantify the effect of flow oscillations on dryout. It is seen that dryout occurs prematurely for certain range of oscillation frequency and amplitude. The experiments which have been carried out till now indicate that better models are required to predict the observed behaviour. Typical OICHF data flow oscillations CHIL facility for OICHF Experiments (OICHF) A mechanistically based model for OICHF is one of the objectives of the phase-III project (b) New reactor designs from a safety systems perspective, passivity and grace time Scope to be discussed further during the meeting

Visits as a part of the collaboration Ms. V. Archana, HBNI to University of Sheffield for a period of 6 months (Jul. 14-Dec. 14). Mr. Arnab Dasgupta, BARC to Imperial College London for a period of 3 months (Nov. 14-Jan. 15). Ms. Parul Goel, HBNI to Imperial College London for a period of 6 months (Dec. 14-May 15). Prof. S.P. Walker and his team have made many short visits to BARC for the experimental collaboration.

Research Output (papers) Development and validation of a model for predicting direct heat transfer from the fuel to droplets in the post dryout regime, Arnab Dasgupta, D K Chandraker, A.K. Nayak, P.K. Vijayan, A. Rama Rao, S.P. Walker, ANS Embedded Topical meeting on Advances in Thermal Hydraulics 2016, New Orleans, LA, [2016] Visualization of large waves in churn and annular two-phase flow, Arnab Dasgupta, D.K. Chandraker, A.K. Nayak, P.K. Vijayan, S.P. Walker, Paper no. 589, Proc. 23rd National and 1st International ASTFE-ISHMT HMTC , [2015] Validation of the Dryout Modelling Code, FIDOM, Dinesh Kumar Chandraker, Arnab Dasgupta, A. K. Nayak, P. K. Vijayan, Kaushik Deshpande, S. P. Walker , Proc. NURETH-16, Chicago, USA, [2015] Quenching of a heated rod: physical phenomena and heat transfer, Arnab Dasgupta, P. P. Kulkarni, G.J. Gorade, D.K. Chandraker, A.K. Nayak, P.K. Vijayan, paper no. 13918, Proc. NURETH-16, Chicago, USA, [2015] Validation and cross verification of three mechanistic codes for annular two-phase flow simulation and dryout prediction, L. Sanmiguel Gimeno, S. P. Walker, G. F. Hewitt, J.-M. Le Corre, A. Dasgupta, M. Ahmad, paper no. 13529, Proc. NURETH-16, Chicago, USA, [2015] Experimental investigation on dominant waves in upward air-water two-phase flow in churn and annular regime Arnab Dasgupta, D.K Chandraker, Suhasith Kshirasagar, B. Raghavendra Reddy, R. Rajalakshmi, A.K. Nayak, S.P. Walker, P.K. Vijayan, G.F. Hewitt, submitted to Experimental Thermal and Fluid Science F.Sebilleau, A.K.Kansal, R.I.Issa, N.K.Maheshwari and S.P.Walker, CFD and experimental analysis of single phase buoyancy driven counter-current flow in a pipe, NURETH-16, Chicago, IL, August 30-September 4, 2015 Frederic Sebilleau, Anuj K. Kansal, Raad I. Issa, Simon P. Walker, “CFD Analysis of Single Phase Counter-Current Buoyancy Driven Flows and its Applications to Passive Reactor Design”, Proceedings of the 22nd International Conference on Nuclear Engineering, ICONE22, July 7-11, Prague, Czech Republic [2014]. An Assessment of the Correlations for Entrainment and Deposition Rates in Annular Flow for Dryout Prediction Arnab Dasgupta, D.K. Chandraker, S.P. Walker, P.K. Vijayan, submitted to Multiphase Science and Technology Measurement of film flow rate and estimation of dryout power in annular flow, Arnab Dasgupta, et al., submitted to Proceedings of the 6th International and 43rd National Conference on Fluid Mechanics and Fluid Power December 15-17, 2016, MNNITA, Allahabad, U.P., India Phenomenological modelling of critical heat flux: The GRAMP code and its validation, M. Ahmad, D.K. Chandraker, G.F. Hewitt, P. K. Vijayan, S.P. Walker, Nuclear Engg. and Design, vol. 254, 280-290, Jan. 2013.

Collaborative Projects on Reactor Physics Collaborators : S.F. Ashley , W.J. Nuttall (Univ. of Cambridge / Open Univ.) and G.T. Parks (Dept. of Engineering, Univ. of Cambridge) An assessment of different open cycle thorium reactors under design were studied by Dr. Stephen Ashley, Cambridge University. Three candidate designs covering LWRs, HWRs and HTGR were chosen namely, AREVA’s European Pressurised Reactor (EPR), Indian Advanced Heavy Water Reactor (AHWR-LEU) General Atomics’ Gas-Turbine Modular Helium Reactor (GT-MHR) It was possible for three Th-U fuelled technologies to be compared to a reference U-fuelled benchmark using the results of modified nuclear reactor models coupled to the UK National Nuclear Laboratory (NNL) code ORION. BARC/RPDD’s contribution was to provide inputs for the AHWR open cycle case study. Prof. William Nuttall (earlier in Cambridge University) and Dr. Stephen Ashley were hosted in RPDD during Dec 2011.

Summary of the Indian participation in the Project (EP/I018425/1) The major conclusions of the Cambridge study are : In terms of the material flows, Th fuelled systems require more separative work units per kWh than the U-fuelled benchmark. This is predominantly due to the requirement of uranium enriched to ∼20% (AHWR requires 6% lesser ore per kWh compared to other reactors). In the analysis it is noted that the fuel fabrication costs for the Th–U-fuelled EPR seed and blanket fuel pins and the AHWR could be underestimated due to the need for novel fabrication techniques and variations in the ratio of UO2 and ThO2 required. In terms of proliferation resistance, Th-LEU fuelled AHWR scores over other reactor designs. Countries will have to choose the option of open cycle with thorium fuel w.r.t to their needs. Comparative performance of the different systems w.r.t sustainability an dproliferation O + ++ -

Research Output (papers) S.F. Ashley , B.A. Lindley, G.T. Parks, W.J. Nuttall, R. Gregg, K.W. Hesketh, U. Kannan, P.D. Krishnani, B. Singh, A. Thakur, M. Cowpere, A. Talamo, “Fuel Cycle Modelling of Open Cycle Thorium-Fuelled Nuclear Energy Systems”, Annals of Nuclear Energy, 69 (2014) pp. 314-330.

Collaborative Projects on Reactor Structures –Fracture mechanics Collaborating Institutes India Reactor Safety Division, BARC UK Prof. R. Ainsworth, Manchester University Project title - Transferability of small-specimen data to large-scale component fracture assessment Comparison of crack initiation points of R6 estimation and fracture expt. Objectives: Validation of R6 methodology by using real life piping experimental data. To determine material J-R curve using R6 failure assessment diagram using experimental data. To assess importance of displacement- based over load-based approach while determining fracture property using R6 methodology.

Facilities Used Pressuring pump Pipe under Four Point Bending

Collaborative Projects on Reactor Structures –Fracture mechanics Conclusions: Estimated crack initiation loads are in good agreement with experimental values for straight pipes. One new simple approach based on R6 is proposed to calculate fracture property using load based approach. This load based approach is further improved by newly proposed displacement based approach.

Summary RDDG, BARC and UK have been collaborating in number of research areas of mutual interest. Several joint publications have arisen from this collaborative work. Researchers of both countries have visited respective labs and gained exposure in relevant areas. Advanced reactor designs require test data and models for T/H, Physics and Structural Mechanics. Data generated in different labs were exchanged for model validation and design. Thus the collaboration is found to be promising and fruitful in this aspect.

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