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A 10 MWe pressurized water reactor for safe and reliable power
65
Nuclear Engineering and Design 109 (1988) 65-71 North-Holland, Amsterdam
A 10 M W e P R E S S U R I Z E D
WATER REACTOR
FOR SAFE AND RELIABLE
POWER
W.H. ARNOLD Westinghouse, Richlana~ Washington, USA R.M. VIJUK, W.R. SHINGLER
and M.C. GROSS
Westinghouse, Madison, Pennsylvania, USA Received April 1988
A design for an innovative, passively safe 10 MWe power plant based on the proven pressurized water reactor technology has been developed. The plant incorporates an innovative design approach to achieve "walk-away" safety and includes significant simplification and elimination of systems and components when compared to the current generation commercial nuclear power plants. The plant has been designed such that the majority of the equipment will be pre-assembled as modules at off-site facilities and shipped to the site on trucks for installation. This approach will provide shorter construction schedules and improved quality control
1. Plant description 1.1. Overall The power plant, fig. 1, consists of an underground power block connected by a service tunnel to a plant
c~~
~Plent a~z,.~,.,ceBuilding ~ ~ ~ "' - "~ ~.
service building located at-grade. The site occupies only 1.4 acres. Key plant parameters are provided in table 1. The power block is contained in a concrete, rectangular structure located underground to enhance plant security and to minimize its presence. The power block interior dimensions are 109-ft long by 57-ft wide vary-
CoolingTowers ~Turbine-Generat°r-
~
Control Room
Power Block
~
"
\\
Containment
Steam Generator ~-;
~
~
Fig. 1. Unobtrusive, small plant requires only 1.4 acres. 0 0 2 9 - 5 4 9 3 / 8 8 / $ 0 3 . 5 0 © E l s e v i e r S c i e n c e P u b l i s h e r s B.V. (North-Holland Physics Publishing Division)
Reactor Vessel
66
W.H. Arnold et al. / A 10 MWe P WR for safe and reliable power
Table 1 Key plant parameters
The above-grade plant service building includes fire protection and water treatment systems, a warehouse,
Overall parameters Site dimension Power block exterior dimension length width depth Operating personnel
115 ft 63 ft 47-67 ft 3 per shift
Plant rating Thermal Electric gross Electric, net Net efficiency Availability factor Design life Plant heat rate
45 MW 12.6 MW 11.5 MW 25.6% - 93% 30 years 12,300 Btu/kW-hr
Core information Fuel type Fuel U-235 enrichment
Fuel assembly lattice Number of assemblies Active core length Core equivalent diameter Fuel loading Refueling interval Refueling time Refueling method Reactor coolant system Working fluid Reactor vessel inlet temp. Reactor vessel outlet temp. Flow rate Pressure, design/operation Secondary system Working fluid Steam generator feedwater temp. Steam generator steam temp. Steam generator pressure, design/operating Flow rate Turbine throttle conditions Condenser backpressure
275 ft by 225 ft
UO 2 3.5% (standard commercial) 17 x 17 (standard commercial) 21 48 in. 43.8 i n . 2.97 MTU 3 years (based on 90% capacity) 14-18 days Under water water 540 o F 561.5 o F 5,695,000 l b / h r 2200 psig/1985 psia water 250 ° F 490 ° F
shops, offices, a n d security. Below g r o u n d are the waste disposal systems, a diesel generator, a technical support center, and switchyard. The s t a n d a r d power block produces 10 M W e (11.5 M W e at the design overpower conditions). By building multiple power blocks at a specific site, larger electrical loads can be accommodated. The same design a p p r o a c h used on this p l a n t can be also extrapolated up to at least 100 M W e without modifying the basic p l a n t concept including the passive safety features. 1.2. Reactor Coolant System
The Reactor Coolant System (RCS), illustrated in fig. 2, consists of a reactor vessel that contains the core, a steam generator for extraction of heat from the primary coolant a n d generation of steam, two p u m p s that pro-
I mproved Steam Generator Integrated Head Package with Water Cooled Roller Nut Control Rod Drive Mechanism ~
IL,'
f
Ii
2200 psig/625 psia 156,000 i b / h r 600 psia/486 o 2.5 in. Hg
"
Reactor
ing in d e p t h from 62 ft to 41 ft. It houses the reactor c o n t a i n m e n t , the fuel storage area (including storage for o n e new core a n d two used cores), the auxiliary a n d control area, a n d the turbine generator area.
~ Canned Motor Pumps
Vessel
Fig. 2. Integral steam generator pump permits close-coupled system.
67
W.H. Arnold et al. / A 10 M W e P W R for safe and reliable power
vide forced coolant circulation, a pressurizer to control system pressure, and associated piping. The RCS has been designed for very low pressure drop to reduce pumping power and to enhance the natural circulation capability of the reactor. Specifically, the RCS has about 40-45% of the pressure drop typical of large commercial PWR plants and can potentially remove 15-20% power using natural circulation. Additional conservatism to improve reliability has been introduced in selection of the thermal hydraulic conditions at which the RCS operates. Since high thermal efficiency is not a primary objective in this design, the reactor vessel outlet coolant temperature is 5 0 - 6 0 ° F lower than that of typical commercial PWR plants. The system employs a "one and one-half" loop configuration that has one vertical U-tube steam generator and employs two canned motor reactor coolant pumps integral with the steam generator. The use of two pumps provides a safety margin to accommodate offnormal events such as a locked pump rotor event. The one and one-half loop configuration has been used in earlier plants, such as the BR-3 in Mol, Belgium that has operated successfully since 1962. While much smaller in size, the steam generator has many features of the Westinghouse Model F design that has been successfully operating in plants since 1980. Material selection, design features, and operating conditions have been selected to assure maximum long-term reliability and serviceability. A Westinghouse Model M-1000 hermetically sealed, canned motor pump has been selected for the plant. This canned motor pump is a low maintenance, high reliability component that requires no seal replacement, thus minimizing downtime. Westinghouse has supplied more than 200 canned motor pumps to 29 utilities for boiler circulation service. The first Westinghouse boiler circulation pump was produced in 1955 and that pump remains in service today. The canned motor pumps have shown high availability factors in both fossil plant boiler service and nuclear plant applications.
The pressurizer and its associated components utilize existing PWR technology. For this application, the size of the pressurizer, in relation to the total RCS volume, has been increased by approximately 25% to minimize the need to add or remove RCS inventory for load changes and cooldown or heatup operations. The larger relative size provides a more conservative response to anticipated transients and does not require actuation of the pressurizer power-operated relief (PORV) or safety valves for anticipated transients. This improves both plant safety and operation. The reactor auxiliary systems consist of commercially available heat exchangers, demineralizers, filters, pumps, tanks, and associated piping, valves, and instrumentation. These systems are simplified versions of the systems used in commercial PWRs. Table 2 indicates that the plant has about 15% of the number of valves and pumps used in large commercial plants. 1.3. Core
The core consists of 21 fuel assemblies, which are a shortened version of the standard Westinghouse 17 x 17 fuel assembly utilizing low enrichment UO 2 with a U-235 concentration of approximately 3.5%. This fuel assembly was selected to take advantage of an experience base of over ten years of successful commercial operation. The overall design of the core is extremely conservative with respect to power density, which is about one-third that used in commercial power plants. The fuel cycle lifetime is three years between refuelings. The core design is simplified compared to a standard PWR in that a single fuel enrichment is used and, with full core refueling, shuffling of fuel assemblies is not required. The core is controlled entirely by control rods. Burnable poison rods permit performance requirements to be achieved without soluble boron in the coolant. The elimination of soluble boron systems for normal operation simplifies the auxiliary systems and improves
Table 2 The plant has simpler nuclear systems than a typical PWR System
Typical PWR Pumps
Component cooling water systems Chemical and volume control systems (Boron recycle) Waste processing system Safety systems
MTP Valves
Pumps
Valves
4 10 13 10
575 590 500 390
2 2 5 0
80 89 93 53
37
2055
9
315
I
TotEs
68
14/.H. A mold et al. / A 10 M We P WR for safe and reliable power
availability compared to commercial PWRs. By increasing the number of reactivity control clusters to thirteen, sufficient margin is provided to achieve cold shutdown, normally performed with the aid of soluble boron. 1.4. P o w e r conversion
The power conversion system arrangement provides a balanced tradeoff between efficiency, availability, and operating simplicity. Saturated steam is delivered to a two stage, single shaft turbine at 600 psia, expanded, and discharged into an all-welded, water cooled condenser. Feedwater is returned to the steam generator via one regenerative low-pressure feedwater heater and one deaerator feedwater heater. The turbine generator and auxiliaries are existing designs that have been in operation in this size power plant. Heat rejected from the steam cycle to the condenser is removed by circulating water and is discharged to the atmosphere by conventional mechanical draft wet cooling modules.
~STI
!
1.5. Electrical s y s t e m
The major electrical equipment and systems are similar to those used on existing power plants. They have been selected and arranged to provide high reliability and ease of maintenance. However, some significant simplifications, including eliminating safety grade diesel generators, have been incorporated. The maximum voltage required within the plant is 480 V. A non-safety grade diesel generator is provided to meet the requirements of plant startup with no external AC power (black start). In addition to serving as a startup source, it can supply backup power in the event off-site power is not available. The plant is provided with two DC systems. One DC system is classified as Class 1E and supplies power to safety related instruments, actuation devices, and reactor protection channels. The second DC system is nonClass IE and supplies power to the breaker control, turbine auxiliaries, computer, and other non-safety related instrumentation.
2. Enhanced safety 2.1. S y s t e m description
The primary side safeguards system (PSSS), shown in fig. 3, forms the basis for achieving a true "walkaway-safe" plant design. This reduces the risk to health and safety as well as the financial risk to the owner. The
Fig. 3. Innovative approach to decay heat removal.
PSSS assures uninterrupted core heat removal and containment integrity following any postulated event. The PSSS makes maximum use of natural phenomena (natural circulation, gravity injection) and fail-to-safe position valves. No pumps, fans, diesels for AC power or operator actions are required to accomplish any safety function. This safety approach has resulted in a significant reduction in the number of safety grade, Class 1E components/systems/structures as compared to current commercial reactor plants. These safeguard features permit system simplification, reduced operator requirements, high reliability, and reduced maintenance, plant size, and plant cost. During normal operation, the two PSSS core makeup tanks (CMTs) are full of "cold" water and are maintained at RCS pressure. The 4-in. CMT discharge piping flowpath from the bottom of the CMT to the reactor vessel is blocked by two normally closed, fail open, parallel isolation valves. Two normally open flowpaths, one 2-in. line from the pressurizer steam space and one 4-in. line from the top of the RCS cold leg, are connected to the top of each CMT. The passive residual heat removal (PRHR) heat exchanger is full of "cold" water and is at RCS pressure. This heat exchanger is submerged in the in-containment water storage tank
W.H. Arnold et al. / A 10 MWe P W R for safe and reliable power
(ICWST) water. The 3-in. line from the PRHR heat exchanger to the RCS cold leg is blocked by two normally closed, fail open, parallel flow control valves. The 3-in. line from the RCS hot leg to the top of the P R H R heat exchanger is normally open. The ICWST is open at the top to the containment and is maintained full of water during normal operation. The single isolation valve in each of the two discharge lines from the ICWST to the reactor vessel are normally open, and the RCS pressure boundary is maintained by two check valves. The containment spray tank (CST), which is located outside containment, is maintained approximately half full of water with a nitrogen gas overpressure. 2.2. Plant trip
Following a plant trip, no immediate actuations of the PSSS functions occur as long as RCS heat removal (via the steam generator) and RCS inventory (pressurizer level) are maintained. If the steam generator is not available for heat removal, the RCS hot leg temperature will increase which will result in a signal to automatically open normally closed valves in the line from the P R H R heat exchanger to the RCS cold leg. The PRHR heat exchanger will transfer heat from the RCS to the water contained in the ICWST, either by natural circulation of RCS fluid from the RCS hot leg to the RCS cold legs, or by forced circulation (reactor coolant pump running) from the RCS cold legs to the hot leg. The ICWST contains sufficient water to prevent steaming from the ICWST to the containment for at least four hours. During this time, restoration of decay heat removal via the steam generator, or initiation of heat removal using the cooldown function of the spent fuel pit cooling system will restore the plant to the normal cooldown mode. If these systems are not or cannot be made available, continued steaming from the ICWST to the containment will ensure RCS heat removal for days, and initiation of the containment ultimate heat sink will assure continued heat removal for an unlimited length of time. Should normal chemical and volume control system RCS makeup not be available following plant trip the CMTs will automatically ensure that RCS inventory is maintained. The relative elevation of CMT (full of cold water) and pressurizer are such that the water in the CM T will flow into the reactor vessel and restore the pressurizer water level. Steam from the pressurizer steam space flows to the top of the CMT maintaining equal pressure in both of the CMTs and pressurizer.
69
2.3. Loss-of-coolant accident
Following a postulated loss-of-coolant accident, the RCS pressure decreases and initiates a reactor trip and safety injection signal which results in a reactor coolant pump trip. As RCS inventory continues to be lost out of the break, the 4-inch piping connection from the top of each RCS cold leg to its corresponding CMT is uncovered, permitting saturated steam to enter the top of the CMT. The elevation difference between the cold water in the CMT and the water level in the corresponding cold leg causes the CMT water to be injected into the reactor vessel. Each CMT can provide an initial injection of 42 l b / s of water which decreases as the CMT empties to 23 lb/s. When the level in either of the CMT decreases to its low level setpoint, the pressurizer PORVs are automatically opened, assuring depressurization of the RCS. Depressurization of the RCS initiates the draining of the ICWST into the reactor vessel. RCS depressurization is assisted by condensation of steam in the P R H R heat exchanger when and if the water level in the RCS is below the top of the loop piping. Draining of the ICWST establishes the long term cooling mode, where eventually, the ICWST water has drained through the reactor vessel ( a n d / o r break) and the water level in the containment building is above the RCS loop piping. In this long-term cooling mode, the core is covered and steams to the containment via the break a n d / o r the pressurizer PORVs. This steam is condensed on the containment steel shell which is cooled via the ultimate heat sink. The water is returned to the reactor vessel via the ICWST a n d / o r directly through the break.
3. Availability The plant design addresses the significant causes of plant outage as identified in the existing data bases. Design differences between this plant and large commercial plants include reduced plant size, less frequent refuelings, increased operating margins, reduced system complexity, and reduced number of components. Using a conservative availability model and the actual experience data, which are adjusted to account for design differences, the availability is predicted to be greater than 90 percent. 3.1. Design improvements
Reduction of system complexity and number of components enhances the availability of the MTP be-
70
W.H. Arnold et al. / A 10 M W e P WR for safe and reliable power
yond that generally experienced in commercial operating reactor plants. For example, the elimination of the boron chemical shim control reduces the number of valves, pipes, and controllers. This reduction in components minimizes forced outages and power reductions experienced by commercial plants. Component reduction, associated with the elimination of the multiple heat transfer loops which are typically found in larger commercial plants, also enhances availability. Component failures are generally accelerated through stress producing mechanisms. The increased operating margins (e.g., coolant temperatures are 5 0 - 6 0 ° F lower than in commercial plants) tend to reduce the component stress producing mechanisms. Increased availability is also a result of the increased redundancy level provided in the reactor trip system. The integrated protection and control scheme uses a two-out-of-four coincidence logic, reverting to a twoout-of-three coincidence logic during test and maintenance intervals. With this operating scheme, the possi-
bility of a single component failure of spurious signal causing a plant trip is eliminated. The modular design and factory fabrication of the MTP will provide additional availability enhancement. The increase quality control and reduced variability of construction factors, such as the use of shop labor as opposed to field labor, minimizes the likelihood of failures. The core design with the low power densities permit a refueling interval of at least three years. With the small core, the refueling outage is estimated to be less than 18 days. An assessment of the maintenance requirements indicates all plant maintenance can be performed in parallel. The steam generator availability is assured by design improvements, the smaller size, and lower operating temperatures. Design improvements, such as the use of thermally treated Inconel 690 tubes, sludge control, all volatile treatment feedwater chemistry, allowance for hydraulic expansion of the steam generator tubes,
\ \
\ \
\
i \, \
./
/
\\~\
\
i! ....
l
.~
/J ~
Fig. 4. The modular design of the plant enhances reliability and reproducibility, and improves construction schedule.
W.H. ArnoM et al. / A 10 M W e P W R for safe and reliable power
71
elimination of copper on the secondary side, and the use of broached, stainless steel support plates, will improve the steam generator reliability. The smaller size and reduced number of steam generator tubes provides additional improvement. Because thermal efficiency is not a primary factor in this design, the steam generator temperatures can be reduced thereby decreasing the potential for corrosion effects. Earlier commercial plants which operated at comparable conditions, such as BR-3 and Yankee Rowe, still use their original steam generators. The canned motor reactor coolant pumps utilized are significantly different from the pumps currently installed in commercial PWRs. Many of the failures associated with reactor coolant pumps are the result of seal leakage brought about by failures within the continuous high pressure seal injection system. The canned motor pump eliminates the need for such a seal and experience indicates a mean time to failure of seven years with the mean time to repair of only 48 h. Smaller sizes and simpler systems within the balance-of-plant also resulted in improvements in availability.
construction site to a manufacturing facility. This reduces manhours at the construction site and reduces the overall schedule since more activities can be performed in parallel. It also allows the work to be performed under the controlled environs of a manufacturing facility, thereby increasing productivity, reducing cost and improving quality. Subsequent identical units will be assembled by the same process and personnel, thereby benefiting from the learning process and further increasing productivity. Modules can be tested in the manufacturing shop, thereby eliminating potential construction rework. Forty-five potential modules have been identified for the plant. Modularization, fig. 4, will be used primarily outside the containment area. Each module is selected to include a functional grouping of equipment thereby permitting interconnections between equipment to be made on the module and minimizing the number of required field interconnections. Modules are generally complete with all hardware including structure, equipment, piping, valves, support, motor control and instrumentation and control cabinets, internal electrical wiring, thermal insulation and finishes.
4. M o d u l a r c o n s t r u c t i o n
5. C o n c l u s i o n
The plant has been designed to minimize site-related activities. The plant has been arranged so that most components can be shipped to a separate factory, away from the site, and assembled into modules which will then be shipped to the site. The modules will have maximum dimensions of 12-feet wide by 12-feet high by 70-feet long. This will permit road shipment throughout the United States. If barge shipment is possible, modules may be further combined at the factory. The combination of shop fabrication of modules, in parallel with site construction of the power block and service buildings, provides a unitized phased construction approach. Modularization transfers labor manhours from the
The 10 MWe plant is the result of applying an innovative design approach to proven pressurized water reactor technology. Key features include: - compact arrangement of proven components, - added margins with respect to core design and plant operating temperatures, - simple systems throughout NSSS and BOP, - passive safety, - modular construction. The plant design results in significant improvements, when compared to existing commercial units, with regard to reliability, operability, plant safety, and plant construction time.
Nuclear Engineering and Design 109 (1988) 65-71 North-Holland, Amsterdam
A 10 M W e P R E S S U R I Z E D
WATER REACTOR
FOR SAFE AND RELIABLE
POWER
W.H. ARNOLD Westinghouse, Richlana~ Washington, USA R.M. VIJUK, W.R. SHINGLER
and M.C. GROSS
Westinghouse, Madison, Pennsylvania, USA Received April 1988
A design for an innovative, passively safe 10 MWe power plant based on the proven pressurized water reactor technology has been developed. The plant incorporates an innovative design approach to achieve "walk-away" safety and includes significant simplification and elimination of systems and components when compared to the current generation commercial nuclear power plants. The plant has been designed such that the majority of the equipment will be pre-assembled as modules at off-site facilities and shipped to the site on trucks for installation. This approach will provide shorter construction schedules and improved quality control
1. Plant description 1.1. Overall The power plant, fig. 1, consists of an underground power block connected by a service tunnel to a plant
c~~
~Plent a~z,.~,.,ceBuilding ~ ~ ~ "' - "~ ~.
service building located at-grade. The site occupies only 1.4 acres. Key plant parameters are provided in table 1. The power block is contained in a concrete, rectangular structure located underground to enhance plant security and to minimize its presence. The power block interior dimensions are 109-ft long by 57-ft wide vary-
CoolingTowers ~Turbine-Generat°r-
~
Control Room
Power Block
~
"
\\
Containment
Steam Generator ~-;
~
~
Fig. 1. Unobtrusive, small plant requires only 1.4 acres. 0 0 2 9 - 5 4 9 3 / 8 8 / $ 0 3 . 5 0 © E l s e v i e r S c i e n c e P u b l i s h e r s B.V. (North-Holland Physics Publishing Division)
Reactor Vessel
66
W.H. Arnold et al. / A 10 MWe P WR for safe and reliable power
Table 1 Key plant parameters
The above-grade plant service building includes fire protection and water treatment systems, a warehouse,
Overall parameters Site dimension Power block exterior dimension length width depth Operating personnel
115 ft 63 ft 47-67 ft 3 per shift
Plant rating Thermal Electric gross Electric, net Net efficiency Availability factor Design life Plant heat rate
45 MW 12.6 MW 11.5 MW 25.6% - 93% 30 years 12,300 Btu/kW-hr
Core information Fuel type Fuel U-235 enrichment
Fuel assembly lattice Number of assemblies Active core length Core equivalent diameter Fuel loading Refueling interval Refueling time Refueling method Reactor coolant system Working fluid Reactor vessel inlet temp. Reactor vessel outlet temp. Flow rate Pressure, design/operation Secondary system Working fluid Steam generator feedwater temp. Steam generator steam temp. Steam generator pressure, design/operating Flow rate Turbine throttle conditions Condenser backpressure
275 ft by 225 ft
UO 2 3.5% (standard commercial) 17 x 17 (standard commercial) 21 48 in. 43.8 i n . 2.97 MTU 3 years (based on 90% capacity) 14-18 days Under water water 540 o F 561.5 o F 5,695,000 l b / h r 2200 psig/1985 psia water 250 ° F 490 ° F
shops, offices, a n d security. Below g r o u n d are the waste disposal systems, a diesel generator, a technical support center, and switchyard. The s t a n d a r d power block produces 10 M W e (11.5 M W e at the design overpower conditions). By building multiple power blocks at a specific site, larger electrical loads can be accommodated. The same design a p p r o a c h used on this p l a n t can be also extrapolated up to at least 100 M W e without modifying the basic p l a n t concept including the passive safety features. 1.2. Reactor Coolant System
The Reactor Coolant System (RCS), illustrated in fig. 2, consists of a reactor vessel that contains the core, a steam generator for extraction of heat from the primary coolant a n d generation of steam, two p u m p s that pro-
I mproved Steam Generator Integrated Head Package with Water Cooled Roller Nut Control Rod Drive Mechanism ~
IL,'
f
Ii
2200 psig/625 psia 156,000 i b / h r 600 psia/486 o 2.5 in. Hg
"
Reactor
ing in d e p t h from 62 ft to 41 ft. It houses the reactor c o n t a i n m e n t , the fuel storage area (including storage for o n e new core a n d two used cores), the auxiliary a n d control area, a n d the turbine generator area.
~ Canned Motor Pumps
Vessel
Fig. 2. Integral steam generator pump permits close-coupled system.
67
W.H. Arnold et al. / A 10 M W e P W R for safe and reliable power
vide forced coolant circulation, a pressurizer to control system pressure, and associated piping. The RCS has been designed for very low pressure drop to reduce pumping power and to enhance the natural circulation capability of the reactor. Specifically, the RCS has about 40-45% of the pressure drop typical of large commercial PWR plants and can potentially remove 15-20% power using natural circulation. Additional conservatism to improve reliability has been introduced in selection of the thermal hydraulic conditions at which the RCS operates. Since high thermal efficiency is not a primary objective in this design, the reactor vessel outlet coolant temperature is 5 0 - 6 0 ° F lower than that of typical commercial PWR plants. The system employs a "one and one-half" loop configuration that has one vertical U-tube steam generator and employs two canned motor reactor coolant pumps integral with the steam generator. The use of two pumps provides a safety margin to accommodate offnormal events such as a locked pump rotor event. The one and one-half loop configuration has been used in earlier plants, such as the BR-3 in Mol, Belgium that has operated successfully since 1962. While much smaller in size, the steam generator has many features of the Westinghouse Model F design that has been successfully operating in plants since 1980. Material selection, design features, and operating conditions have been selected to assure maximum long-term reliability and serviceability. A Westinghouse Model M-1000 hermetically sealed, canned motor pump has been selected for the plant. This canned motor pump is a low maintenance, high reliability component that requires no seal replacement, thus minimizing downtime. Westinghouse has supplied more than 200 canned motor pumps to 29 utilities for boiler circulation service. The first Westinghouse boiler circulation pump was produced in 1955 and that pump remains in service today. The canned motor pumps have shown high availability factors in both fossil plant boiler service and nuclear plant applications.
The pressurizer and its associated components utilize existing PWR technology. For this application, the size of the pressurizer, in relation to the total RCS volume, has been increased by approximately 25% to minimize the need to add or remove RCS inventory for load changes and cooldown or heatup operations. The larger relative size provides a more conservative response to anticipated transients and does not require actuation of the pressurizer power-operated relief (PORV) or safety valves for anticipated transients. This improves both plant safety and operation. The reactor auxiliary systems consist of commercially available heat exchangers, demineralizers, filters, pumps, tanks, and associated piping, valves, and instrumentation. These systems are simplified versions of the systems used in commercial PWRs. Table 2 indicates that the plant has about 15% of the number of valves and pumps used in large commercial plants. 1.3. Core
The core consists of 21 fuel assemblies, which are a shortened version of the standard Westinghouse 17 x 17 fuel assembly utilizing low enrichment UO 2 with a U-235 concentration of approximately 3.5%. This fuel assembly was selected to take advantage of an experience base of over ten years of successful commercial operation. The overall design of the core is extremely conservative with respect to power density, which is about one-third that used in commercial power plants. The fuel cycle lifetime is three years between refuelings. The core design is simplified compared to a standard PWR in that a single fuel enrichment is used and, with full core refueling, shuffling of fuel assemblies is not required. The core is controlled entirely by control rods. Burnable poison rods permit performance requirements to be achieved without soluble boron in the coolant. The elimination of soluble boron systems for normal operation simplifies the auxiliary systems and improves
Table 2 The plant has simpler nuclear systems than a typical PWR System
Typical PWR Pumps
Component cooling water systems Chemical and volume control systems (Boron recycle) Waste processing system Safety systems
MTP Valves
Pumps
Valves
4 10 13 10
575 590 500 390
2 2 5 0
80 89 93 53
37
2055
9
315
I
TotEs
68
14/.H. A mold et al. / A 10 M We P WR for safe and reliable power
availability compared to commercial PWRs. By increasing the number of reactivity control clusters to thirteen, sufficient margin is provided to achieve cold shutdown, normally performed with the aid of soluble boron. 1.4. P o w e r conversion
The power conversion system arrangement provides a balanced tradeoff between efficiency, availability, and operating simplicity. Saturated steam is delivered to a two stage, single shaft turbine at 600 psia, expanded, and discharged into an all-welded, water cooled condenser. Feedwater is returned to the steam generator via one regenerative low-pressure feedwater heater and one deaerator feedwater heater. The turbine generator and auxiliaries are existing designs that have been in operation in this size power plant. Heat rejected from the steam cycle to the condenser is removed by circulating water and is discharged to the atmosphere by conventional mechanical draft wet cooling modules.
~STI
!
1.5. Electrical s y s t e m
The major electrical equipment and systems are similar to those used on existing power plants. They have been selected and arranged to provide high reliability and ease of maintenance. However, some significant simplifications, including eliminating safety grade diesel generators, have been incorporated. The maximum voltage required within the plant is 480 V. A non-safety grade diesel generator is provided to meet the requirements of plant startup with no external AC power (black start). In addition to serving as a startup source, it can supply backup power in the event off-site power is not available. The plant is provided with two DC systems. One DC system is classified as Class 1E and supplies power to safety related instruments, actuation devices, and reactor protection channels. The second DC system is nonClass IE and supplies power to the breaker control, turbine auxiliaries, computer, and other non-safety related instrumentation.
2. Enhanced safety 2.1. S y s t e m description
The primary side safeguards system (PSSS), shown in fig. 3, forms the basis for achieving a true "walkaway-safe" plant design. This reduces the risk to health and safety as well as the financial risk to the owner. The
Fig. 3. Innovative approach to decay heat removal.
PSSS assures uninterrupted core heat removal and containment integrity following any postulated event. The PSSS makes maximum use of natural phenomena (natural circulation, gravity injection) and fail-to-safe position valves. No pumps, fans, diesels for AC power or operator actions are required to accomplish any safety function. This safety approach has resulted in a significant reduction in the number of safety grade, Class 1E components/systems/structures as compared to current commercial reactor plants. These safeguard features permit system simplification, reduced operator requirements, high reliability, and reduced maintenance, plant size, and plant cost. During normal operation, the two PSSS core makeup tanks (CMTs) are full of "cold" water and are maintained at RCS pressure. The 4-in. CMT discharge piping flowpath from the bottom of the CMT to the reactor vessel is blocked by two normally closed, fail open, parallel isolation valves. Two normally open flowpaths, one 2-in. line from the pressurizer steam space and one 4-in. line from the top of the RCS cold leg, are connected to the top of each CMT. The passive residual heat removal (PRHR) heat exchanger is full of "cold" water and is at RCS pressure. This heat exchanger is submerged in the in-containment water storage tank
W.H. Arnold et al. / A 10 MWe P W R for safe and reliable power
(ICWST) water. The 3-in. line from the PRHR heat exchanger to the RCS cold leg is blocked by two normally closed, fail open, parallel flow control valves. The 3-in. line from the RCS hot leg to the top of the P R H R heat exchanger is normally open. The ICWST is open at the top to the containment and is maintained full of water during normal operation. The single isolation valve in each of the two discharge lines from the ICWST to the reactor vessel are normally open, and the RCS pressure boundary is maintained by two check valves. The containment spray tank (CST), which is located outside containment, is maintained approximately half full of water with a nitrogen gas overpressure. 2.2. Plant trip
Following a plant trip, no immediate actuations of the PSSS functions occur as long as RCS heat removal (via the steam generator) and RCS inventory (pressurizer level) are maintained. If the steam generator is not available for heat removal, the RCS hot leg temperature will increase which will result in a signal to automatically open normally closed valves in the line from the P R H R heat exchanger to the RCS cold leg. The PRHR heat exchanger will transfer heat from the RCS to the water contained in the ICWST, either by natural circulation of RCS fluid from the RCS hot leg to the RCS cold legs, or by forced circulation (reactor coolant pump running) from the RCS cold legs to the hot leg. The ICWST contains sufficient water to prevent steaming from the ICWST to the containment for at least four hours. During this time, restoration of decay heat removal via the steam generator, or initiation of heat removal using the cooldown function of the spent fuel pit cooling system will restore the plant to the normal cooldown mode. If these systems are not or cannot be made available, continued steaming from the ICWST to the containment will ensure RCS heat removal for days, and initiation of the containment ultimate heat sink will assure continued heat removal for an unlimited length of time. Should normal chemical and volume control system RCS makeup not be available following plant trip the CMTs will automatically ensure that RCS inventory is maintained. The relative elevation of CMT (full of cold water) and pressurizer are such that the water in the CM T will flow into the reactor vessel and restore the pressurizer water level. Steam from the pressurizer steam space flows to the top of the CMT maintaining equal pressure in both of the CMTs and pressurizer.
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2.3. Loss-of-coolant accident
Following a postulated loss-of-coolant accident, the RCS pressure decreases and initiates a reactor trip and safety injection signal which results in a reactor coolant pump trip. As RCS inventory continues to be lost out of the break, the 4-inch piping connection from the top of each RCS cold leg to its corresponding CMT is uncovered, permitting saturated steam to enter the top of the CMT. The elevation difference between the cold water in the CMT and the water level in the corresponding cold leg causes the CMT water to be injected into the reactor vessel. Each CMT can provide an initial injection of 42 l b / s of water which decreases as the CMT empties to 23 lb/s. When the level in either of the CMT decreases to its low level setpoint, the pressurizer PORVs are automatically opened, assuring depressurization of the RCS. Depressurization of the RCS initiates the draining of the ICWST into the reactor vessel. RCS depressurization is assisted by condensation of steam in the P R H R heat exchanger when and if the water level in the RCS is below the top of the loop piping. Draining of the ICWST establishes the long term cooling mode, where eventually, the ICWST water has drained through the reactor vessel ( a n d / o r break) and the water level in the containment building is above the RCS loop piping. In this long-term cooling mode, the core is covered and steams to the containment via the break a n d / o r the pressurizer PORVs. This steam is condensed on the containment steel shell which is cooled via the ultimate heat sink. The water is returned to the reactor vessel via the ICWST a n d / o r directly through the break.
3. Availability The plant design addresses the significant causes of plant outage as identified in the existing data bases. Design differences between this plant and large commercial plants include reduced plant size, less frequent refuelings, increased operating margins, reduced system complexity, and reduced number of components. Using a conservative availability model and the actual experience data, which are adjusted to account for design differences, the availability is predicted to be greater than 90 percent. 3.1. Design improvements
Reduction of system complexity and number of components enhances the availability of the MTP be-
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W.H. Arnold et al. / A 10 M W e P WR for safe and reliable power
yond that generally experienced in commercial operating reactor plants. For example, the elimination of the boron chemical shim control reduces the number of valves, pipes, and controllers. This reduction in components minimizes forced outages and power reductions experienced by commercial plants. Component reduction, associated with the elimination of the multiple heat transfer loops which are typically found in larger commercial plants, also enhances availability. Component failures are generally accelerated through stress producing mechanisms. The increased operating margins (e.g., coolant temperatures are 5 0 - 6 0 ° F lower than in commercial plants) tend to reduce the component stress producing mechanisms. Increased availability is also a result of the increased redundancy level provided in the reactor trip system. The integrated protection and control scheme uses a two-out-of-four coincidence logic, reverting to a twoout-of-three coincidence logic during test and maintenance intervals. With this operating scheme, the possi-
bility of a single component failure of spurious signal causing a plant trip is eliminated. The modular design and factory fabrication of the MTP will provide additional availability enhancement. The increase quality control and reduced variability of construction factors, such as the use of shop labor as opposed to field labor, minimizes the likelihood of failures. The core design with the low power densities permit a refueling interval of at least three years. With the small core, the refueling outage is estimated to be less than 18 days. An assessment of the maintenance requirements indicates all plant maintenance can be performed in parallel. The steam generator availability is assured by design improvements, the smaller size, and lower operating temperatures. Design improvements, such as the use of thermally treated Inconel 690 tubes, sludge control, all volatile treatment feedwater chemistry, allowance for hydraulic expansion of the steam generator tubes,
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Fig. 4. The modular design of the plant enhances reliability and reproducibility, and improves construction schedule.
W.H. ArnoM et al. / A 10 M W e P W R for safe and reliable power
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elimination of copper on the secondary side, and the use of broached, stainless steel support plates, will improve the steam generator reliability. The smaller size and reduced number of steam generator tubes provides additional improvement. Because thermal efficiency is not a primary factor in this design, the steam generator temperatures can be reduced thereby decreasing the potential for corrosion effects. Earlier commercial plants which operated at comparable conditions, such as BR-3 and Yankee Rowe, still use their original steam generators. The canned motor reactor coolant pumps utilized are significantly different from the pumps currently installed in commercial PWRs. Many of the failures associated with reactor coolant pumps are the result of seal leakage brought about by failures within the continuous high pressure seal injection system. The canned motor pump eliminates the need for such a seal and experience indicates a mean time to failure of seven years with the mean time to repair of only 48 h. Smaller sizes and simpler systems within the balance-of-plant also resulted in improvements in availability.
construction site to a manufacturing facility. This reduces manhours at the construction site and reduces the overall schedule since more activities can be performed in parallel. It also allows the work to be performed under the controlled environs of a manufacturing facility, thereby increasing productivity, reducing cost and improving quality. Subsequent identical units will be assembled by the same process and personnel, thereby benefiting from the learning process and further increasing productivity. Modules can be tested in the manufacturing shop, thereby eliminating potential construction rework. Forty-five potential modules have been identified for the plant. Modularization, fig. 4, will be used primarily outside the containment area. Each module is selected to include a functional grouping of equipment thereby permitting interconnections between equipment to be made on the module and minimizing the number of required field interconnections. Modules are generally complete with all hardware including structure, equipment, piping, valves, support, motor control and instrumentation and control cabinets, internal electrical wiring, thermal insulation and finishes.
4. M o d u l a r c o n s t r u c t i o n
5. C o n c l u s i o n
The plant has been designed to minimize site-related activities. The plant has been arranged so that most components can be shipped to a separate factory, away from the site, and assembled into modules which will then be shipped to the site. The modules will have maximum dimensions of 12-feet wide by 12-feet high by 70-feet long. This will permit road shipment throughout the United States. If barge shipment is possible, modules may be further combined at the factory. The combination of shop fabrication of modules, in parallel with site construction of the power block and service buildings, provides a unitized phased construction approach. Modularization transfers labor manhours from the
The 10 MWe plant is the result of applying an innovative design approach to proven pressurized water reactor technology. Key features include: - compact arrangement of proven components, - added margins with respect to core design and plant operating temperatures, - simple systems throughout NSSS and BOP, - passive safety, - modular construction. The plant design results in significant improvements, when compared to existing commercial units, with regard to reliability, operability, plant safety, and plant construction time.