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
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
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.
Idaho National Engineering and Environmental Laboratory
ANS Classification of Events
• Classification of Accident events per ANSI N18.21973 (industry standard based on ANS committee)
Condition I:
Condition II:
Condition III:
Condition IV:
Normal operation and operational transients
Faults of moderate frequency
Infrequent faults
Limiting faults
• Classification according to expected frequency of
occurrence
• Less frequent events may have more severe
consequences
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
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)
Idaho National Engineering and Environmental Laboratory
Transient temperature limit met
when AFW flow exceeded 15%
• 5-s MFW
coastdown
10% AFW
20% AFW
30% AFW
800
• Scram
o
Temperature ( C)
1000
• Constant
pressure
600
400
200
0
10
20
Time (s)
30
40
Idaho National Engineering and Environmental Laboratory
Temperature limit met for 50%
step change in MFW flow
1200
• No scram
• No AFW
o
Temperature ( C)
1000
800
25% step
50% step
75% step
100% step
600
400
0
10
20
Time (s)
30
40
Idaho National Engineering and Environmental Laboratory
Fast-opening 100%-capacity turbine
bypass system helps significantly
1000
• 5-s MFW
coastdown
o
Temperature ( C)
900
800
• Scram
700
• No AFW
Constant pressure
20% steam relief
100% steam relief
600
500
400
0
10
20
Time (s)
30
40
Idaho National Engineering and Environmental Laboratory
Higher coolant density reactivity
feedback lowers cladding temperature
1000
o
Temperature ( C)
• 5-s MFW coastdown
• Scram
800
• No AFW
600
400
Low feedback
High feedback
0
10
20
Time (s)
30
40
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%)
Idaho National Engineering and Environmental Laboratory
Pressure response following a
turbine trip is acceptable
Pressure (MPa)
30
0% SRV, no scram
90% SRV, no scram
80% SRV, scram
28
• Instant control
valve closure
• Continued MFW
at rated flow
26
24
0
2
4
Time (s)
6
8
10
Idaho National Engineering and Environmental Laboratory
Small increase in reactor power
following turbine trip
1.5
Normalized power
• Instant control
valve closure
1.0
• Continued MFW
0% SRV, no scram
90% SRV, no scram
80% SRV, scram
0.5
0.0
0
2
4
Time (s)
6
8
10
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