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Idaho National Engineering and Environmental Laboratory SCWR Preliminary Safety Considerations Cliff Davis, Jacopo Buongiorno, INEEL Luca Oriani, Westinghouse.

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Presentation on theme: "Idaho National Engineering and Environmental Laboratory SCWR Preliminary Safety Considerations Cliff Davis, Jacopo Buongiorno, INEEL Luca Oriani, Westinghouse."— Presentation transcript:

1 Idaho National Engineering and Environmental Laboratory SCWR Preliminary Safety Considerations Cliff Davis, Jacopo Buongiorno, INEEL Luca Oriani, Westinghouse Electric Co. April 29, 2003 Madison, Wisconsin

2 Idaho National Engineering and Environmental Laboratory Introduction Safety concept and classification of the events Parametric thermal-hydraulic calculations of the SCWR during loss-of-feedwater and turbine-trip transients to determine the required response time and capacities of safety systems Calculations used the RELAP5 computer code, which has been recently improved for SCWR applications Analysis was performed for a design with solid moderator rods, but the results are expected to be more generally applicable Transient cladding temperature limit of 840  C was used to evaluate the thermal-hydraulic response

3 Idaho National Engineering and Environmental Laboratory Safety Concept Active, non-safety systems have passive, safety-related back-up to perform nuclear safety functions –Safety functions automatically actuated, no reliance on operator action –Passive features actuated by stored energy (batteries, compressed air) –Once actuated, their continued operation relies only on natural forces (gravity, natural circulation) with no motors, fans, diesels, etc. Common approach with the most advanced LWR concept proposed by the main NSSS vendors: –Westinghouse AP600/AP1000, IRIS and System 80+ –Framatome-ANP SWR-1000 –GE ESBWR and ABWR Design Goal: Achieve a degree of safety at least comparable to the more advanced plant concepts being currently proposed.

4 Idaho National Engineering and Environmental Laboratory ANS Classification of Events Classification of Accident events per ANSI N (industry standard based on ANS committee) Condition I: Normal operation and operational transients Condition II: Faults of moderate frequency Condition III: Infrequent faults Condition IV: Limiting faults Classification according to expected frequency of occurrence Less frequent events may have more severe consequences

5 Idaho National Engineering and Environmental Laboratory The loss-of-feedwater and turbine-trip transients were evaluated because SCWR is a once-through direct cycle without coolant recirculation in the reactor vessel –Loss of feedwater is important because It results in rapid undercooling of the core It is a Condition II event that must not result in any significant damage to the fuel Average coolant density is low in the SCWR core and pressurization events result in significant positive reactivity insertion –Turbine trip without steam bypass has the potential to cause a significant increase in reactor power

6 Idaho National Engineering and Environmental Laboratory Parametric calculations for loss of feedwater investigated the effects of Main feedwater (MFW) coastdown time (0 to 10 s) Scram (with and without) Auxiliary feedwater (AFW) flow rate (10-30% of rated feedwater) Steam relief (20-100% capacity) Step changes in MFW flow rate (25-100%) Coolant density reactivity feedback (nominal and high)

7 Idaho National Engineering and Environmental Laboratory Transient temperature limit met when AFW flow exceeded 15% 5-s MFW coastdown Scram Constant pressure

8 Idaho National Engineering and Environmental Laboratory Temperature limit met for 50% step change in MFW flow No scram No AFW

9 Idaho National Engineering and Environmental Laboratory Fast-opening 100%-capacity turbine bypass system helps significantly 5-s MFW coastdown Scram No AFW

10 Idaho National Engineering and Environmental Laboratory Higher coolant density reactivity feedback lowers cladding temperature 5-s MFW coastdown Scram No AFW

11 Idaho National Engineering and Environmental Laboratory Parametric calculations of a turbine trip without steam bypass investigated the effects of Scram Safety relief valve (SRV) capacity (0 - 90%)

12 Idaho National Engineering and Environmental Laboratory Pressure response following a turbine trip is acceptable Instant control valve closure Continued MFW at rated flow

13 Idaho National Engineering and Environmental Laboratory Small increase in reactor power following turbine trip Instant control valve closure Continued MFW

14 Idaho National Engineering and Environmental Laboratory Conclusions SCWR with solid moderator rods can tolerate a 50% step change in MFW flow without scram Transient temperature limit can be met following a total loss of MFW if AFW flow exceeds 15% of initial MFW flow AFW flow requirements can be reduced by –Fast-opening 100%-capacity turbine bypass –Higher feedback coefficients typical of designs with water rods Acceptable pressure response following turbine trip without steam bypass if the SRV capacity is greater than 90% Power increase following turbine trip without steam bypass and with full MFW flow is much smaller than in comparable BWRs


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