Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

Transcription

1 Idaho National Engineering and Environmental Laboratory Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Nuclear Energy Research Initiative Project INEEL Principal Investigators: Philip MacDonald Jacopo Buongiorno Cliff Davis Kevan Weaver Collaborating Organizations: Massachusetts Institute of Technology University of Michigan Westinghouse Electric Company

2 Objectives Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design (INEEL). For the fast-spectrum SCWR, metallic (dispersion type), and oxide fertile fuels will be investigated to evaluate the void and Doppler reactivity coefficients, actinide burn rate, and reactivity swing throughout the irradiation cycle. For the thermal-spectrum SCWR, a variety of fuel and moderator types will be assessed. Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking (University of Michigan and MIT). MIT will use an existing supercritical-water loop to conduct corrosion experiments in flowing supercritical water. A high temperature autoclave containing a mechanical test device will be built in Year 1 and operated in Years 2 and 3 at the University of Michigan to collect stress-corrosion cracking data. The data from both universities will be used to identify promising structural and fuel cladding materials and develop appropriate corrosion and stress corrosion cracking correlations.

3 Objectives (Cont.) Task 3. Plant Engineering and Reactor Safety Analysis (Westinghouse and INEEL). The optimal configuration of the power conversion cycle will be identified. Particular emphasis will be given to the applicability of current supercritical fossil-fired plant technology and experience to a direct-cycle nuclear system. A steady-state sub-channel analysis of the reactor core will be undertaken with the goal of establishing power limits and safety margins under normal operating conditions. Also, the reactor susceptibility to coupled neutronic/thermalhydraulic oscillations will be evaluated. The response of the plant to accident situations and anticipated transients without scram will also be assessed.

4 Fast Reactor Neutronics Long Fuel Cycles and Relatively Low Reactivity Swing are Possible in Fast Spectrum SCWRs ThN (20% TRU) UN (20% TRU) UN (15.3% TRU) k-inf MWd/kg 204 MWd/kg 223 MWd/kg Effective Full Power Years k-inf ThZr (20% TRU) UZr (20% TRU) UZr (13.1% TRU) 140 MWd/kg 196 MWd/kg 242 MWd/kg Effective Full Power Years

5 Thermal Reactor Neutronics k-eff Alloy 718 penalization is significant (arbitrary units) H2O ZrH1.6 Be BeO C SiC Zr-4 Alloy 718 ZrH1.6, H2O, PWR SCWR Be, C PWR - Be and C fail to generate a significant thermal tail in the neutron spectrum -ZrH x and water rods perform similarly 0 1.0E E E E E E+01 E (MeV)

6 Thermal Reactor Neutronics - Local Peaking - Low local peaking: The hot pin is (10,10) - Peaking can be reduced by decreasing enrichment in the corner pins (red), and increasing it in the intersection pins (green) Pin (10,10) Pin (1,1)

7 Key issues analyzed for the ZrH x solid moderator: - Thermal and Irradiation Stability - ZrH x Thermal Capacity - ZrH x / Water Interaction - Hydrogen Release and Redistribution - ZrH x / Alloy 718 compatibility - Fabrication and Costs

8 University of Michigan Supercritical Water Loop System

9 304L Sample Tested at 550 C, 25.4Mpa, 8ppm O 2, and 10-7 %Elongation/sec About 20 cracks/mm 2 appear along the entire length of the tensile bar. The same material tested under PWR and BWR conditions did not exhibit this type of surface cracking.

10 304L Fracture Surface Analysis Intergranular fracture Ductile fracture

11 MIT Sample Rack and Exposure Cell

12 Pitting of Alloy Hours at 400 C in DI Water

13 Pitting of Alloy Hours at 400 C in DI Water

14 Design Criteria for SCWRs Maximum Allowable Cladding Temperature The principal fuel system design criterion for SCWR analyses is defined in terms of a maximum allowable cladding temperature (MAT): MAT-I shall be defined as the limit temperature at which the cladding can successfully operate for the required core lifetime. MAT-II shall be defined as the cladding temperature for which there is a 95% probability with a 95% confidence that significant fuel rod damage or failure will not occur during the limited duration of any design basis accident. Initial limits for the SCWR for Alloy 718: MAT-I = 620 C MAT-II = 840 C

15 The Large Core Enthalpy Rise and the Pseudotransition of the SCW Phase Make the SCWR Very Sensitive to Variations in the Main T&H Parameters Core Enthalpy Rise [KJ/kg] Open Lattice, No Orificing Results 0 PWR - AP1000 BWR - ABWR SCWR A 10% variation in flow or power in a PWR will lead to an increase in the core outlet temperature of only 2-4 C. The same variation in a SCWR will cause an increase in the outlet temperature of about C. Temperature [C] Nominal Channel Hot Channel Distance from Core Inlet (fraction of total core length)

16 Summary A qualitative analysis was performed to determine which fuel form would support the highest reactivity-limited burnup in a fast-spectrum SCWR, and would have proliferation resistant isotopics. The best fuel appears to be a mixture of thorium and uranium to balance long core life with proliferation resistant isotopics. The neutronic performance of several solid moderators for use in a thermal spectrum SCWR core was evaluated and compared to that of water rods. The only acceptable solid moderator is delta-phase zirconium hydride (ZrH 1.6 ), which generates a relatively high multiplication factor and a negative coolant void reactivity coefficient. Several issues key to the chemical and thermo-mechanical feasibility of ZrH 1.6 were assessed including zirconium-hydride/water interaction, hydrogen release, hydrogen redistribution, pressurization of the moderator box at high temperature, phase stability, and compatibility of zirconium hydride with the moderator box material. The design and fabrication of the University of Michigan supercritical water loop system for stress corrosion cracking tests was completed and experiments have begun.

17 Summary (Cont.) The initial test results indicate that type 304L stainless steel, which is commonly used in LWRs, is highly susceptible to stress corrosion cracking in SCWRs. Initial experiments over a temperature range encompassing both suband supercritical conditions have been completed at MIT with 316L and I-625 samples. Localized breakdown and surface pitting both for exposed and occluded regions was observed. A preliminary core layout (dimensions, core configuration) and thermal-hydraulic design (temperature, pressure, flow rates) was developed. The design criteria for the system has been defined and the correlations for heat transfer in supercritical water, and the methods for the hot channel factors that will be used to verify the proposed design criteria, have been identified. An analysis of the temperatures and density profiles in the average and hot channels for different possible system configurations was completed. A completely satisfactory design has not yet been identified.