Development Of The System Safety Analysis Code For Critical And External Neutron Driven Sub-critical Fast-spectrum Reactor

With the development of nuclear energy,the disposal of spent fuel is becoming more and more serious.Accelerator driven sub-critical system(ADS)can effectively reduce the long-life,high-level radioactive nuclides in spent fuel.Internationally,many advanced countries in nuclear energy have paid much attention to and supported the research of ADS.In China,the new concept the Accelerator Driven Advanced Nuclear Energy System(ADANES)has been proposed by the Institute of Modern Physics(IMP)on the basis of ADS.Moreover,the domestic ADS research has been transferred from concept research to integrated system construction and the ultimate goal is to build high-power industrial devices.On the support of ADS project,developing and testing the fast-spectrum reactor system safety analysis code with independent intellectual property rights can not only help to provide effective ways for safety analysis and solution optimization of the subcritical reactor,but also lay a foundation for the construction of industrial demonstration equipment.What`s more,it can enrich the project results and promote the development of the domestic system safety analysis code.In this paper,firstly,the fast-spectrum reactor NE-TH coupling safety analysis code IMPC-transient was developed.The main work includes the independent development of the thermal-hydraulics code and the development of the NE-TH coupling interface between the deterministic analysis code DAISY.A reasonable and complete thermal-hydraulic model of the primary loop is established in the code,including the core model,the heat exchanger model,the pool model,the pipe model,the pump model,etc.The equation is discrete using the finite volume method,and the algebraic equations are solved by Gaussian elimination or Jacobian point iteration method.The code is developed with Fortran and the thermal-hydraulic code and the neutronic code DAISY are coupled in source-level.At present,the code can be operated in both pure thermal and NE-TH coupling modes.Moreover,in the neutronics model,either the zero-dimensional point kinetics equation,or the three-dimensional transport kinetics equation is available.Secondly,the benchmark of EBR-? shutdown heat removal tests SHRT-17 and SHRT-45 R were used to verify IMPC-transient.The predicted results were compared with the measured results as well as the results predicted by other organizations.The result shows that the results predicted by IMPC-transient are mostly in good agreement with the measured results.Though a little of predicted results are different from the measured results,the error is in the same level compared with other organizations.It proves that IMPC-transient is correct and reliable in both modes.And then,IMPC-transient was used in the safety analysis of the 7.5MW lead-bismuth cooled subcritical reactor.A 3D NE-TH calculation model was established and five conditions were simulated including the steady condition,protected loss of flow(PLOF),unprotected loss of flow(UPLOF),unprotected beam overpower(UBOP)and Beam-trip(BT).The result shows that under the steady condition,the distribution of key parameters such as neutron flux,power density,flow rate and temperature are reasonable,which indicates that the steady design of the reactor is reasonable.Moreover,under these four transient conditions,the maximum temperature of the fuel,cladding and coolant are within the safety limits,and the temperature difference of the cladding under the one second Beam-trip condition is small,indicating that the design is safe under these four transient conditions.Finally,in order to study the heat transfer between the spallation target and the reactor,a rapid analysis code was developed to calculate the temperature of the target region.With the help of this code,the temperature field of the 2.5MW spallation target was simulated.In this code,a theoretical model for the rapid calculation of the target temperature is proposed based on the Lagrange algorithm.The code was developed in MATLAB,which was then verified by Fluent.The result shows that the temperature field calculated by the code is in good agreement with Fluent.Moreover,the heat deposited in the target can be removed safely,and the heat transfer between the target and the reactor can be neglected.Also,it shows that the axial heat conduction doesn`t have much effect on the temperature distribution.