Design and Performance of South Ukraine NPP Mixed Cores with Westinghouse Fuel

Transcription

1 National Science Center "Kharkov Institute of Physics and Technology (NSC KIPT) Design and Performance of South Ukraine NPP Mixed Cores with Westinghouse Fuel A.M. Abdullayev, V.Z. Baidulin, A.I. Zhukov Nuclear Fuel Cycle Science and Technology Establishment (NFC STE) Center for Reactor Core Design (CRCD) Ukraine

2 Contents 2 Ukraine Fuel Project Goals Safety and Licensing Issues for Mixed Core Nuclear Design Methods Qualification Summary of Mixed Core Operation with WFAs Mixed Core Transition and Equilibrium Cycles Design Safety Analysis for Mixed Cores Conclusion

3 Ukraine Fuel Qualification Project 3 Qualify alternative nuclear fuel supply in Ukraine Establish an organization in Ukraine to receive fuel and core design and analysis methodologies Implement and operate Westinghouse fuel assemblies (WFAs) in Ukraine's NPPs

4 Ukraine Fuel Qualification Project 4 CRCD was established in CRCD staff ~25 people with experience in nuclear fuel design and operation (scientists, computer specialists, NPP specialists). 8 PhDs and Doctors of science. 12 specialists were trained during 5 years at Westinghouse. Design tools methodology, codes, equipment.

5 Safety and Licensing Issues 5 Mixed Core Challenges: Non-uniform hydraulics Different engineering safety factors for power peaking coefficients A lack of experience in mixed core analysis methods in Ukraine Necessity to qualify the Westinghouse core design methods for resident fuel TVSA TVSM WFA

6 Safety and Licensing Issues (cont d) 6 Mixed Core Requirements: Maintain SUNPP core loading strategy Do not change the operating algorithms and the setpoints of the control and protection systems Maintain core design limits and safe operation conditions established for the cores with resident fuel Meet all safety limits for the mixed core with current reactor system design

7 7 Safety and Licensing Issues (cont d) Regulatory framework Requirements of Ukraine s regulatory and technical documents (General NPP Safety Rules OPB AES- 2008, etc.) for nuclear and radiation safety Requirements for Content of Safety Analysis Report for NPPs with WWER Reactors operating in Ukraine, Regulatory Document RD-95 SNRCU Management order #65 Requirements for new fuel implementation, 2002 Requirements for Modification of Nuclear Facilities and Procedure for Their Safety Evaluations, NP

8 8 Safety and Licensing Issues (cont d) The main documents submitted to the regulatory authorities: Qualification reports on analysis methodologies used for FA design substantiation and WWER-1000 safety analysis ( ) Technical task document for WFA design (2003) WFA mechanical and hydraulic test reports (2004) WFA WWER-1000 mechanical design report (2010) Technical Specifications for WFA ( ) Safety substantiation reports for the use of WFAs at SUNPP-2 and SUNPP-3 ( )

9 9 Safety and Licensing Issues (cont d) About 120 technical reports on safety substantiation and support of WFAs have been submitted to the regulatory authority

10 Mixed Core Engineering 10 Mixed Core Solutions: Qualify nuclear design calculation tools against operational data and other codes Use standard and improved transient analysis methods (e.g. 3D neutron kinetics for RIA analysis) Use conservative engineering safety factors for the WFA

11 Computer Codes Qualification 11 Comparison of Predicted and Measured Data Comparison of predicted nuclear design characteristics and measured data against plant data. Comparison of predicted fuel assembly (FA) peaking factors with Monte Carlo calculations performed using the MCNPX code. The difference between the predicted and the measured core neutronic parameters are within the uncertainties of calculation methodologies

12 Critical Boron Concentration SUNPP-3 Cycles (Prediction vs. Measurement) 12 Критическая Critical Boron концентрация Concentration, бора, g/kg г/кг Critical Критич. Boron бор (эксплуат. (operational данные) data) Crit. Критич. Boron бор (ANC-H расчет) prediction) Critical Boron, g/kg Power and Bank 10 Position, % Cb measured Power Bank EFPDs Эфф. сут. Burnup, EFPDs

13 Critical Boron Concentration 13 SUNPP-2 Cycles (Prediction vs. Measurement) SUNPP-2 Cycle 21 SUNPP-2 Cycle 22 Operational data Operational data EFPDs EFPDs

14 HZP Measurements 14 (Prediction vs. Measurement) Boron Concentration Isothermal Coefficient Lead Bank Worth Trip Reactivity

15 ZNPP-3 Cycle 20 Peaking Factors (Prediction vs. Measurement) SUNPP-2 Cycle ZNPP-3 Cycle 20 Axial Nodal Peaking Factor Kv SUNPP-2 Cycle 20 FA Relative Power Kq

16 Pin Factor (PHOENIX-H vs MCNPX) 16 0 MWD/MТU MWD/MТU Burnup Range MWD/MТU) Pin Factor MSD (PHOENIX-H - MCNPX) = 0.008

17 ANC-H/PHOENIX-H Qualification 17 Parameter Designation Unit Critical boron concentration (at HZP) Moderator temperature coefficient (fuel+moderator) Boron reactivity coefficient Lead Bank worth Trip reactivity Lead Bank differential worth Mean square deviation Based on Operational Data σ = 0.04 for the predicted and measured axial nodal peaking factors (К v ) σ = 0.02 for the predicted and measured assembly relative power (K q ) σ = for the fuel rod relative power for predicted ANC-H/PHOENIX-H and MCNPX and accounting for burnup (K r ) C B cr g/kg 0.24 ρ/ t 10-3 %/ С 0.94 ρ/ C %/g/kg 0.11 ρ Lead Bank % 0.06 ρ trip reactivity % 0.61 ρ Lead Bank / Н 10-3 %/cm 0.29

18 6 LTAs Operation Summary 18 Cycle Max K Q Max F h Max burnup, T eff, Operation year MW D/tU EFPDs

19 42 WFAs Operation Summary 19 Inspection Type WFA drag force Drag force during WFA loading into the core WFA top nozzle axial difference RCCA drop time RCCA drag force during loading into/withdrawal from the WFA Leakage test WFA visual inspection Results Drag force did not exceed 50 kgf. One FA had the maximum drag force of 80 kgf Drag force did not exceed 75 kgf. For a number of FAs in the central zone the drag force reached 150 kgf Did not exceed 5 mm after the first operating cycle Maximum control rod drop time throughout cycle 2.1 sec Within specified limits, not exceeding 4 kgf No leaking fuel rods were identified No damage was identified preventing further WFA operation Unit 3 Reload Batch operation - Cycle 21 (270.3 EFPDs)

20 Transition Core Design 20 Transition core design accounts for: Assurance of core power requirements Nuclear compatibility and interchangeability with the current TVSA inventory Meeting design restrictions for all jointly operated fuel types Power reduction in the hot WFA channel due to the hydraulic irregularity of the core

21 SUNPP-3 Fuel Cycles Cycle 21 Cycle 22

22 Core Peaking Factors for SUNPP Max WFA FDH cycle Max WFA FDH cycle FDH FDH 1.35 H10=70% H10=90% 1.35 H10=70% H10=90% EFFPD EFFPD Cycle 21 Cycle 22

23 23 Equilibrium Fuel Cycle (42 Feed WFAs) Core Power Distribution and FA Average Burnup for the Equilibrium Cycle (BOC) Kq Kr BU Kq Kr BU

24 12-Month Cycle Summary 24 Parameter WFA Equilibrium cycle TVSA Number of FAs per feed, pcs UO 2 weight in FA, kg Average enrichment of feed fuel; U 235 w/o Cycle length, EFPDs Maximum burnup of unloaded FAs MW D/tU Maximum Kq Maximum Kr Maximum Ko Reactivity coefficients ρ/ t m at HZP, BOC, [10-5 / C] ρ/ γ at HZP, BOC, [%/(g/cm 3 )] ρ/ t U at HZP, BOC, [10-5 / C] Bank 10 worth at HFP ρ Lead Bank BOC, % ρ Lead Bank end of cycle (EOC), % Trip reactivity with the highest worth RCCA withdrawn at HFP ρ trip reactivity. BOC, % ρ trip reactivity EOC, %

25 25 Transition Cores Analysis (SUNPP-2) Utilizing sub-channel analysis for crossflow and DNBR evaluation Cycle Cycle Cycle 26 Cycle Rod ID Channel ID X Индекс твэла YY Индекс канала XX Rod ID YY Channel ID Instrumental Инструментальная труба tube Guide tube Труба направляющего канала hot WFA TVSA TVSM VIPRE 192-channel model

26 Hot Channel Penalty WFA Fuel Rod Power Reduction, % ch. Linear (192-ch.) Fraction WFA in the VVER-1000 core with TVSA

27 Safety Analysis Transients DNB Margin, % TVSA WFA 10 0 Inadvertent opening of BRU-K Feedwater line break TG trip MSIV closure Boron Dilution Inadvertent rod withdrawal Trip of 1 MCP Loss of power on 4 MCPs

28 3D Neutron Kinetics for RIA 28 NESTLE few group neutron diffusion equation solver for steady-state and transient based on the nodal expansion method (burnup gradient and pin power reconstruction) Improved RIA analysis - more realistic but still conservative in comparison with the licensed 1D methods Separate hot channel analysis by VIPRE using power setpoints from NESTLE 3D core calculation (NESTLE) Nuclear design model 3D power shapes feedbacks TH model of average and hot channel Coolant temperature and density Fuel temperature Hot channel calculation (VIPRE) Maximum cladding temperature Maximum fuel temperature Maximum radial averaged enthalpy

29 HFP Rod Ejection Accident 29 Fuel rod design differences require extensive evaluation to demonstrate that fuel related safety criteria are met Parameter 1-D 3-D Maximum fuel temperature, С Maximum cladding temperature, С Radial average peak fuel enthalpy, J/g

30 Conclusion 30 Operation of Westinghouse Lead Test Assemblies was successfully completed Mixed cores with WFAs for WWER-1000 are operated at SUNPP-2,3 successfully Core design methods and codes have been continuously verified based on the operational data. Safety of mixed cores and equilibrium cycles (~300 EFPDs) with a feed of 42 WFAs has been demonstrated and agreed by the Regulator ZNPP-5 is expected to start mixed core operation with WFAs in 2012

31 31 Thank you! Any questions?