Research On Transient Thermal Hydraulic Characteristics Of AP1000 Under Typical Accident Conditions

As a typical Generation ?+ PWR,AP1000 adopts a series of natural forces,including gravity,natural circulation,natural convection and compressed gas swell to ensure its passive safety features in an innovative way.According to the national strategy for development of nuclear power,higher power advanced PWRs such as CAP1400 and CAP1700 with our own independent intellectual property rights will be developed on the basis of introducing,digesting and absorbing AP1000 nuclear power technology.Currently the conceptual design of CAP1400 has been finished basically.In this dissertation,key issues in transient thermal hydraulic characteristics of AP1000 under typical accident conditions were investigated at various levels by adopting different analysis tools.Firstly,a set of resonable and complete mathematical and physical models were established according to the structural and operational characteristics of AP1000,including core model,natural circulation steam generator model,electrically heated pressurizer model,main pump model,passive residual heat removal system(PRHRS)model,critical flow model and auxiliary models.In the non-LOCA transient thermal-hydraulic system code LOFTRAN developed by U.S.Westinghouse for AP600 and AP1000,in terms of the steam generator model,a lumped parameter two-region model in the steam generator secondary side was adopted.But in this dissertation,a more advanced and more resonable distributed parameter model was adopted.As for the pressurizer model,compared with the two-region non-equilibrium model in LOFTRAN,three region and multi region non-equilibrium models were provided for selection.In addition,an innovative PRHRS model was established based on basic mass,momentum and energy conservation equations.Furthermore,a non-LOCA transient thermal-hydraulic system code named RETAC(REactor Transient Analysis Code)was developed using FORTRAN structural language.Modular programing technology was adopted by RETAC and it was convenient for further modification and development of the RETAC code.Based on the mathematical and physical models established,the primary loop system and PRHRS of AP1000 were divided into a large number of control volumes.Gear method was used for solution of ordinary differential equations obtained.The steady state results by RETAC were compared with the rated values in the Westinghouse DCD(Design Control Document)and the comparation showed a good agreement.Moreover,turbine trip accident and ADS(Automatic Depressurization System)inadvertent operation accident were analyzed and the calculation results were compared with results by the large commercial code RELAP5 and the Westinghouse LOFTRAN code respectively.Good agreement proved the rationality and accuracy of the RETAC modeling.Furthermore,the RETAC code was applied to analysis of typical non-LOCA accidents in AP1000,including loss of flow accident(partial loss of flow,complete loss of flow and pump rotor locked),SG feed water temperature reduction accident,PRHRS inadvertent operation accident and ADS inadvertent operation accident(with and without offsite power).The calculation results showed that the MDNBR(Minimum Departure from Nucleate Boiling Ratio)in the core was always higher than the safety analysis limit value 1.5 and met the safety criterion demand.Among them,the calculation results of the PRHRS inadvertent operation accident agreed well with those by the LOFTRAN code developed by Westinghouse and the THEATRe/JTopmeret code developed by GSE and thus the rationality of the PRHRS model was proven.An analysis model for AP1000 primary loop system and passive safety systems,including core makeup tanks(CMTs),accumulators(ACCs),in-containment refueling water storage tank(IRWST),PRHRS and ADS,was established by large commercial code RELAP5/MOD3.4.Several typical small break LOCA transients,including 5.08cm(2 in.),10.16cm(4in.),20.32cm(8 in.)and 25.40cm(10 in.)cold leg small break,were analyzed.The analysis results showed that during small break LOCAs in AP1000,the maximum core void fraction did not exceed the safety analysis limit value ? =0.9 and the dry out CHF(Critical Heat Flux)would not occur.The PCT(Peak Cladding Temprature)was far below the safety analysis limit value 1478K/2200?in Appendix K.It proved that the action of the passive safety systems could remove the core residual heat effectively and thus ensure the reactor safety.Finally,liquid entrainment through ADS stage four(ADS-4)in the late phase of small break LOCA in AP1000 was analyzed.As an important thermal hydraulic phenomenon during small break LOCA transients,liquid entrainment through ADS-4 influences coolant inventory in the primary side and cooling of the primary loop and the core region.In this dissertation,the liquid entrainment subroutine hzflow in the RELAP5/MOD3.0 was modified and compiled by adopting the liquid entrainment onset and quality correlations obtained by the ATLATS test facility at Oregon State University.Further,the modified RELAP5 version was applied to analysis and calculation of 5.08cm(2 in.)typical small break LOCA in AP1000.Differences between the calculation results before and after modification of the RELAP5 code showed that the liquid entrainment model in RELAP5 underestimated the liquid entrainment quantity through ADS-4 and resulted in a non-conservative safety analysis results.The research work in the dissertation is meaningful for construction and safety operation of AP1000 and can offer references for research and development of CAP1400 and CAP1700 in the future.