SCWRs are basically LWRs operating at higher pressure And temperatures with a direct oncethrough
cycle (pic2.). Operation above the critical pressure eliminates coolant boiling, so the coolant
remains single-phase throughout the system.
Thus the need for recirculation And
jet pumps, pressurizer, steam
generators, steam separators And
dryers is eliminated (pic.3). The
main mission of the SCWR is
generation of low-cost electricity.
Thus the SCWR is also suited for
hydrogen generation with
electrolysis, And can support the
development of the hydrogen
economy in the near term. It is
built upon two proven
technologies, LWRs, which are the
most commonly deployed power
generating reactors in the world,
And supercritical fossil-fired boilers, Picture 2. SCWR - simplified once-through direct cycle
a large number of which is also in use around the world. The SCWR concept is being investigated by
32 organizations in 13 countries.
The objective of the multi-year SCWR program is to assess the technical viability of the SCWR
concept. Thus the focus is on establishing a conceptual design, assessing its safety And stability
characteristics, And identifying And testing cAndidate materials for all reactor components.
The US supercritical light water reactor (SCLWR) with a thermal spectrum will be in this report the
subject of the most development work And is the basis for much of the reference design.
The reference design has been selected a for the SCWR system that focuses on a large-size, directcycle,
thermal-spectrum, light-water
cooled And moderated, low-enriched
uranium fuelled, base-load operation
plant for electricity generation at low
capital And operating costs. The
operating pressure And core inlet/outlet
temperatures are 25 MPa And
280/500oC, respectively. The coolant
density decreases from about 760
kg/m3 at the core inlet to about 90
kg/m3 at the core outlet. Thus, large
square water rods with down flow are
used to provide adequate moderation
in the core. The fuel pin design is
similar to that of a pressurized water
reactor (PWR), but with higher fill pressure And longer Picture 3. Simplifications in SCWR
fission gas plenum. CAndidate materials for all fuel assembly And vessel internal components include
ferritic-martensitic steels And low-swelling austenitic steels for the components exposed to high
neutron doses, And high-strength austenitic steels And nickel-based alloys for low-dose components.
However, these materials are un-proven in the potentially aggressive SCWR environment, And their
performance will have to be tested. A materials development program has been prepared for this
purpose.
Two traditional austenitic steels (304L And 316L) were tested for corrosion And stress-corrosion
cracking (SCC) susceptibility in supercritical water. It was found that both alloys are susceptible to
SCC (316L less so than 304L) in both
deaerated And nondeaerated hightemperature
(>400oC) supercritical
water. Thus, these alloys cannot be
used for hightemperature components
in the SCWR. However, they could be
used for components operating in the
280-350oC range (e.g., the lower core
plate, the control rod guide tubes),
given their satisfactory behavior in
deaerated water at these temperatures.
The SCWR core average power
density is about 70 kW/L, i.e., between
the power density of boiling water
reactors (BWRs) And PWRs. The reactor Picture 4. Operating cycle diagrams for SCWR
coolant system of the SCWR comprises the feedwater lines And main steam lines up to the outermost
set of containment isolation valves. Similar to a BWR, the SCWR uses two feedwater lines made of
carbon steel. However it has been determined that because of its high-density steam, the SCWR needs
only two steam lines as opposed to four in a BWR of similar thermal power. This further adds to the
economic strength of the SCWR concept. The steam lines can be constructed out of ferritic steels such
as P91 And P92, which are currently used in supercritical fossil plant steam lines.
A pressure-suppression type containment with a condensation pool, essentially the same design as
modern BWRs, was selected. The dry And wet well volumes were calculated to limit the pressure
build-up to typical BWR levels following a LOCA or a severe accident with core melting.
The condensation pool water inventory was designed to provide ample margin for residual heat
removal And meet the requirement that active safety systems are not needed during the first 24 hours
following an initiating event resulting in a severe accident. The very conservative European Utility
Requirements for mitigation of severe accidents were adopted in sizing the containment And a core
catcher was added to the design. Despite this conservative approach the SCWR containment is
somewhat smaller than that of an advanced BWR of similar thermal power, And thus significantly
smaller on a per unit electric power basis.
Thermal-hydraulic And thermal-nuclear coupled instabilities were investigated at ANL with a
frequency-domain linear stability analysis code based on single-channel thermal hydraulics, onedimensional
fuel heat conduction, And point-kinetics models. The BWR stability criteria were adopted
And it was found that the SCWR is stable against core-wide in-phase oscillations at normal operating
power And flow conditions.
A critical review of the LWR abnormal events And their NRC classification has been performed with
the SCWR application in mind. Four events were singled out that could be potentially troublesome: (i)loss of feedwater flow, which in the once-through direct-cycle SCWR coincides with the loss of core
flow, (ii) turbine trip without steam bypass, which pressurizes the system And could result in
significant positive reactivity insertion because of the low density of the SCWR coolant, (iii) loss of
feedwater heating, which also results in the insertion of positive reactivitybecause of the lack of
feedwater mixing with hotter coolant in the vessel, And (iv) large break in the feedwater lines, which,
if unmitigated, results in coolant stagnation in the core And rapid overheating of the fuel. A
preliminary analysis of these four key events was performed at INEEL with a modified version of the
RELAP5 code. It was found that the SCWR behavior is relatively benign during the turbine trip
without steam bypass, the loss of feedwater heating, And the large break in the feedwater lines.
On the other hAnd, survival of the total loss of feedwater will likely require the use of a high-capacity
high-pressure fast-acting auxiliary feedwater system. Design of such system will be a major challenge.
The reference SCWR system has a power conversion cycle that is very similar to a supercritical coalfired
plant, with the boiler replaced by the nuclear reactor. A conceptual study was performed by
BREI to identify an optimal configuration for the goals of thermal efficiency maximization And capital
cost minimization. The SCWR power conversion cycle uses a single-shaft turbine-generator, operating
at reduced speed (1,800 rpm), with one high-pressure/intermediate-pressure (HPT/IPT) turbine unit
And three low-pressure turbine (LPT) units with six flow paths, with a moisture separator reheater
between the HPT/IPT And the LPTs, eight feedwater heaters, steam-turbine-driven feedwater pumps
And natural draft cooling towers. The reference design generates 1,600 MWe with a thermal efficiency
(net electric power to the grid / fission power) of 44.8% versus about 35% for LWRs under equivalent
assumptions.
A pre-conceptual design of the SCWR control system was also performed. The main characteristics
affecting the design of the SCWR control system are the relatively low vessel water inventory, the
nuclear/thermal-hydraulic coupling, the lack of level indication under supercritical conditions And the
absence of recirculation flow. The main variables to be controlled include the reactor power, the core
outlet temperature during supercritical pressure operation (e.g., full power operation), the reactor
pressure, the reactor level during subcritical pressure operation (e.g., during start-up) And the
feedwater flow. Then, assuming base-load operation, the recommended approach for the SCWR is one
in which the control rods accomplish the primary control of the thermal power, the turbine control
valve provides the control of the pressure, the feedwater flow (i.e., the feedwater pumps) provides the
primary control of the outlet temperature, And the control of the coolant inventory in the vessel is
accomplished by assuring that steam And feed flow are balanced while maintaining the correct core
outlet temperature.
Also, rather than an approach in which higher functions such as power or turbine valve control are in
manual with lower level control loops in automatic, the use of an integrated control approach, one in
which all functions are in automatic, is deemed preferable due to the SCWR’s expected fast response
to perturbations.
In summary, the research work during the first year of the Generation-IV SCWR program has
confirmed the basic assumptions contained in the Generation-IV Roadmap Report regarding the
SCWR, And no new potential showstoppers have been found. The key feasibility issues for the SCWR
remain the development of in-core materials And the demonstration of adequate safety. Dynamic
instabilities appear to be less of a concern. |
发表于 2010-9-14 16:19:18