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Nuclear Power: Technical and Institutional Options for the Future (1992)

Chapter: 3 Assessment of Advanced Nuclear Reactor Technologies

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Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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The narratives describing the various technologies are based upon the oral and written record submitted to the Committee. These summaries represent a conscientious effort to accurately depict the nature, attributes, and distinguishing features of each technology. The Committee does not represent this Chapter as a comprehensive treatment of each advanced nuclear reactor technology or as an independent verification of all vendor representations.

3

Assessment of Advanced Nuclear Reactor Technologies

The Committee was asked to perform a critical comparative analysis of the practical technological options for the future development of nuclear power. In conducting this analysis the Committee undertook the following tasks:

  • identifying the full range of practical nuclear reactor technologies for the next generation of nuclear plants;

  • developing criteria to evaluate these technologies; and

  • evaluating the technologies in terms of the criteria developed.

The Committee developed evaluation criteria that reflected the characteristics deemed most important for future nuclear power plants (e.g., safety and cost). (see Appendix B) The Committee then invited reactor vendors to present design concepts for advanced nuclear reactor technologies. Enhanced and novel features of these technologies are first described, and then the technologies are evaluated in light of the Committee's criteria.

OVERVIEW OF ADVANCED REACTOR TECHNOLOGIES

Most reactors operate by fissioning uranium atoms with slow or thermal neutrons. Thermal neutrons are produced in moderators such as graphite or water. The reactor cores are usually cooled with water or a gas (e.g., helium). Some reactors have no moderator, operate with fast neutrons, and are normally cooled by a liquid metal (e.g., sodium). A summary of the advanced reactor technologies reviewed by the Committee is given in Table 3-1, based on vendor-provided information. The major headings in Table 3-1 (Large Evolutionary Light Water Reactors, etc.) align with the titles of the major sections below in which the advanced reactors are discussed. The acronyns in Table 3-1 are explained in the following paragraphs.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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TABLE 3-1 Vendor Descriptions of Technical Aspects of Advanced Nuclear Reactors

Reactor Designationa

Vendor

Power (MWe)

Passive Containment Cooling

Passive Residual Heat Removal

Passive Emergency Core Cooling System

Primary Coolant

Digital Control

Large Evolutionary Light Water Reactors

ABWR

GE

[GE Nuclear Energy, 1989]

1,350

No

No

No

Water

Yes

APWR-1300b

Westinghouse

[McCulchan et. al., 1989]

1,350

No

No

No

Borated water

Yes

System 80+ PWR

CEc

[CE, 1989a]

1,300

Being Evaluated

No

No

Borated water

Yes

Mid-Sized Light Water Reactors With Passive Safety Features

AP-600 PWR

Westinghouse

[Westinghouse, 1989]

615

Yes

Yes

Yes

Borated water

Yes

SBWR

GE

[GE Nuclear Energy, 1989]

600

Yes

Yes

Yes

Water

Yes

Other Reactor Concepts

CANDU 3 HWR

AECL

[AECL, Undated; AECL, 1989]

450

No

Yesi

Yes/Noe

Heavy water

Yes, with automated startup

SIR PWR

CEc

[CE, 1989b]

320h

Yes

Yes

Yes/Noe

Water

Yes

MHTGR

GA

[(GA, 1989]

134f

Yes

Yesd

Yesd

Helium

Yes, with automated startup

PIUS PWR

ABB Atom

[ABB, 1989]

640

Yes

Yesd

Yesd

Borated water

Yes

PRISM LMR

GE

[Berglund, 1989; Till, 1989]

155g

Yes

Yesd

Yesd

Sodium

Yes, with automated startup

a All designs include load-following capability of between 50 and 100 percent.

b Westinghouse also supplied information on the APWR-1000, a 1,050 MWe plant whose features are similar to the APWR-1300, except for the core design.

c Combustion Engineering; now Asca Brown Boveri Combusion Engineering Nuclear Power.

d The residual heat removal system and the emergency core cooling system are essentially the same system.

e Yes at high pressure; no at low pressure.

f The power plant design includes four 134 MWe reactor modules for a total of 536 MWe.

g The power plant design includes one to three power blocks, each containing three 155 MWe reactor modules for a total of 465 MWe per block, and a net electrical plant rating up to 1,395 MWe.

h The power plant design includes two 320 MWe reactor modules on one turbine generator to produce 640 MWe output.

i System under development.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The advanced commercial water reactors reviewed are of three classes: (1) pressurized water reactors (PWR) are light water reactors (LWR) that maintain the water adjacent to the fuel elements at high pressure to prevent boiling; (2) boiling water reactors (BWR) are LWRs in which the water adjacent to the fuel elements boils; and (3) heavy water reactors (HWR) are reactors in which heavy water (deuterium oxide or D2O) serves as both coolant and moderator instead of ordinary (light) water, and only the coolant is pressurized. In current HWRs the reactor fuel is natural uranium, and in LWRs the fuel is uranium enriched to contain up to a few percent of the uranium-235 isotope. (APWR and ABWR mean “advanced”; AP means “advanced passive”; SBWR means “simplified”; CANDU means “Canadian deuterium uranium”; SIR means “safe integral reactor”; and PIUS means “process inherent ultimate safety”.)

Two other advanced reactor technologies reviewed by the Committee do not use water as a coolant or moderator. They are the gas-cooled graphite-moderated reactor known as the MHTGR (modular high-temperature gas-cooled reactor) and the liquid metal-cooled fast neutron reactor known as the PRISM LMR (power reactor, innovative small module liquid metal reactor).

The vendors, in order of appearance in Table 3-1, are General Electric (GE), Westinghouse, Combustion Engineering (CE), Atomic Energy of Canada Limited (AECL), General Atomics (GA), and Asea Brown Boveri (ABB) Atom.

The following sections treat ten advanced reactor types--three large evolutionary LWRs, two mid-sized LWRs with passive safety features, and five other reactor concepts.

Large Evolutionary Light Water Reactors

Evolutionary LWRs, a subset of advanced reactors consisting of the ABWR, APWR-1300, and System 80+, are improved versions of current LWRs with capacities of greater than 1,000 megawatts electric (MWe). These evolutionary designs differ to some extent from current LWRs, for which thousands of reactor years of operating experience have been accumulated worldwide. All evolutionary designs seek greater safety margins, greater ease of construction, improved reliability and availability, improved maintainability, lower costs, and greater ease of operation over existing large LWRs. The evolutionary reactor designs conform to the advanced LWR requirements contained in the Utility Requirements Document.[EPRI, 1990] A summary of these requirements, which cover both enhanced safety and improved economies, is presented in Table 3-2. The Utility Requirements Document is being prepared through the Electric Power Research Institute (EPRI). The technical judgments on all significant issues are reviewed by a Utility Steering Committee made up of experienced nuclear utility executives from throughout the United States and abroad.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

TABLE 3-2 Key Utility Design Requirements for Advanced Light Water Reactors**

Plant size

Reference size 1,200-1,300 MWe for evolutionary designs; reference size 600 MWe for passive safety deigns

Design life

60 years

Design philosophy

Simple, rugged, no prototype required

Accident resistance

≥15 percent fuel thermal margin, increased time for response to upsets

Core damage frequency

< 10−5/year by probabilistic risk analysis

Loss of coolant accident

No fuel damage for 6″ pipe break

Severe accident mitigation

< 25 REM at site boundary for accidents with > 10−6/year cumulative frequency

Emergency planning zone

For passive plant provide technical basis for simplification of off-site emergency plan

Design availability

87 percent

Refueling interval

24 months capability

Maneuvering

Daily load follow

Worker radiation exposure

< 100 person REM/year

Construction time

1,300 MWe: ≤ 54 months (first concrete to commercial operation); 600 MWe: ≤ 42 months

Design status

90 percent complete at construction initiation

Economic goals

10 percent cost advantage over alternatives (nonnuclear) after 10 years and 20 percent advantage after 30 years

Resulting cost goals (1989 $)

Overnight capital 30-year levelized total generation

*1,200 MWe commercial operation in 1998; 600 MWe in 2000

SOURCE: Electric Power Research Institute. Advanced Light Water ReactorUtility Requirements Document, Volume 1, ALWR Policy and Summaryof Top-Tier Requirements. Issued 3/90. Palo Alto, California.

** These requirements apply to both the large evolutionary LWRs and to the mid-sized LWRs with passive safety features.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The first standardized design to be certified in the United States is likely to be an evolutionary LWR. Three of these LWR design concepts were presented to the Committee. Only the new or unique features of each concept will be described.

Advanced Boiling Water Reactor

The 1,350 MWe ABWR is being developed as the next Japanese standard BWR under the leadership of the Tokyo Electric Power Company in a joint venture with GE, Hitachi, Toshiba, and a group of Japanese utilities. In 1989 the Tokyo Electric Power Company announced its decision to proceed with the construction of two ABWR units at its Kashiwazaki-Kariwa Nuclear Power Station, with commercial operation of the first unit scheduled for 1996 and of the second for 1998. GE was selected to supply the nuclear steam supply systems, fuel, and turbine generators. Figure 3-1 is a diagram of this advanced reactor's pressure vessel and core.

Finally, GE has applied for design certification under 10 CFR Part 52, and certification currently is scheduled for completion in the mid-1990s. GE expects that this reactor will be the first certified U.S. standard plant.[Wolfe and Wilkens, 1988]

Core Design. A new core and fuel design has been developed to increase operating economies, and external recirculation pumps have been replaced by internal pumps. The reactor pressure vessel has a single forged ring for the 10 internal pump nozzles and the conical support skirt. The elimination of the external recirculation pump piping and the use of the vessel forged rings have resulted in a 50 percent reduction in the weld requirements for the primary system pressure boundary. Finally, the reactor pressure vessel is standard BWR design, except that (1) the annular space between the pressure vessel shroud and the vessel wall is increased, and (2) the standard cylindrical vessel support is now a conical skirt.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-1 Advanced boiling water reactor, pressure vessel and core.

SOURCE: GE Nuclear Energy

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Fluid Systems. The emergency core cooling system and residual heat removal system have a three-division scheme. Two divisions each provide both high-pressure and low-pressure emergency core coolant injection capability. The third division combines a reactor-steam-driven turbine pump for the high-pressure coolant injection and low-pressure coolant injection system. The steam driven system is the conventional reactor core isolation cooling (RCIC) system that has been upgraded to a safety system. The other two divisions are the high-pressure core flooders. The steam driven system is controlled by water level and is the first high-pressure system to come on in the event of a loss-of-coolant accident (LOCA) or a reactor isolation transient. The residual heat removal system is a triply redundant water delivery/decay heat removal combination. Additionally, the elimination of large nozzles on the reactor vessel below the core helps ensure that the core is not uncovered during any LOCA. At the same time, a 50 percent reduction of the total required emergency core cooling system pumping capacity is realized, compared to an equivalent-size external loop BWR plant.

Control and Instrumentation. The control and instrumentation system features a multiplexing system that complements a digital, solid-state control design. This equipment permits a design that increases the system redundancy, provides fault-tolerant operation, and provides self-diagnostics while the system is in operation. 1

Containment. The reactor building/containment is a steel-lined reinforced concrete structure with a covered pressure suppression pool. The design also features a horizontal vent system for venting the drywell to the suppression pool in the event of a LOCA. In addition, elimination of the external recirculation piping system permits greater access for inspection and maintenance of the drywell.

1  

Multiplexing will be considered, as will all the advanced instrumentation and controls technology, as part of the licensing process for the large evolutionary reactors. This will establish the precedent for the other advanced reactors. Included in the licensing review will be digital controls technology and the new control room designs that incorporate current human factors considerations.[M. Chiramal, Section Chief, Advanced Reactor Section, Instrumentation and Control, U.S. Nuclear Regulatory Commission, personal communication, August 29, 1991.]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×
Advanced Pressurized Water Reactor

The design for a large evolutionary APWR has been developed by Westinghouse in cooperation with five Japanese utilities and Mitsubishi. Kansai Electric Company has declared its intention to build the first such plant, pending approval of a suitable site.[Hirata et al., 1989] This four-loop 1,350 MWe model incorporates several technological advancements.[McCutchan et al., 1989] Although it was primarily developed for Japan, the design concepts were adopted in the criteria specified by EPRI. Figure 3-2 depicts the reactor's integrated safety systems.

Core Design. The most significant new feature of the APWR is the 15 to 20 percent reduction in power density for greater safety and thermal operating margins. Reactivity is controlled with rods that displace water in the lattice during the first part of the refueling cycle; the water is returned later in the cycle by removing the displacement rods. (This feature is not included in the APWR-1000 design, which has a conventional but reduced power density core.) It is claimed that these features combine to reduce fuel costs by 20 percent. In addition, the increase in the number of movable rods compared to conventional designs requires a larger rod-guide region above the core. The larger reactor vessel provides an increased inventory of cooling water above the core, leading to enhanced safety while reducing requirements for the emergency core cooling system (ECCS).

Steam Generators. The U-tube steam generators are larger than those in existing Westinghouse reactors, with lower average temperatures, lower heat flux, and easier accessibility for maintenance and repair. Other features include improved tube materials and an improved tube support plate design.

Fluid Systems. Safety and control functions have been integrated, reducing piping requirements and enhancing safety-related fluid system design. For the ECCS, four high-pressure pumps take suction from an in-containment refueling water storage tank and inject borated water into the reactor vessel to improve core protection for small pipe breaks. This eliminates the switchover from a tank located outside the containment to a sump inside the containment.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-2 Advanced pressurized water reactor integrated safety systems (1 = Accumulator; 2 = High head safety injection pump; 3 = Residual heat removal heat exchanger; 4 = Residual heat removal/coolant systems pump).

SOURCE: Westinghouse Energy Systems

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Control and Instrumentation. The integrated control safety systems feature microprocessors and multiplexed data highways that allow complete and rapid communication between the central control room and the various control and protection points in the plant. The multiplexed interconnections reduce control cabling by up to 70 percent. The safety system is designed to operate automatically when plant conditions reach trip set points.

Containment. A double cylindrical containment building is used with an interior pressure bearing steel shell and an external concrete shield wall. The steel containment shell is easier to construct to quality standards. The total containment volume is increased, and congested areas have been eliminated.

System 80+ Standard Design Pressurized Water Reactor

The System 80+ PWR, the third large evolutionary reactor reviewed by the Committee, is rated at 1,300 MWe. It is the result of a design effort led by CE (now Asea Brown Boveri Combustion Engineering Nuclear Power), assisted by the Duke Power Company and the Korea Advanced Energy Research Institute. This design evolved from CE's System 80 nuclear steam supply system design. The advanced System 80+ design draws heavily on the designs of three operating System 80 units at Palo Verde and two more scheduled for construction in Yonggwang, Republic of Korea. Incremental improvements to the components that are currently used have been incorporated in the new design.[CE, 1989a] Figure 3-3 is an elevation view of the System 80+ containment building.

Core Design. The System 80+ core design uses only control rods for reactivity control, thus eliminating the need to adjust the boron concentration in the coolant. This feature simplifies reactivity control during power load changes. In addition, the core thermal operating margin has been increased by reducing normal operating hot leg temperatures and revising monitoring methods.

Steam Generators. Design enhancements in the steam generators include better steam dryers, an increased overall heat transfer area, and slightly reduced full power steam pressure resulting from lower coolant temperatures, compared to the System 80 design. Additional heat transfer area permits the nuclear steam supply system to maintain rated output with a significant

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-3 Elevation view of System 80+ containment. SOURCE: [CE, 1989a]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

number of tubes plugged. Each steam generator will also have a larger secondary feedwater inventory which extends the “boil dry” time, enhancing the nuclear steam supply system's capability to tolerate upset conditions and thereby improving operational reliability.

Fluid Systems. The safety injection system in the enhanced System 80+ is a four-train system of injection pumps used for both low-pressure and high-pressure injection of borated water into the reactor coolant system. This feature eliminates the requirement for a dedicated low-pressure injection system and associated cross-connects with the shutdown cooling system. In addition, four separate safety injection tanks are part of the safety injection system. The in-containment refueling water storage tank eliminates the reliance on automatic or manual switchover of suction in the event of a break in the primary coolant piping.

Control and Instrumentation. The System 80+ control system features a new design to meet human factor, reliability, and licensing requirements. It is characterized by digital processing, fiber optic data communications, and touch-sensitive video displays.

Containment. The System 80+ containment design is a 200-foot-diameter pressure bearing steel sphere surrounded by an outer concrete shield building. The concrete shield that surrounds the steel sphere offers secondary containment, and the relatively large free internal volume (3.4 million cubic feet) provides increased capacity for absorbing energy and diluting hydrogen concentrations in the event of an accident. Finally, the steel shell acts as a natural heat sink and offers the potential for passive heat removal using external cooling. This steel containment building is designed with an operating floor that offers 75 percent more usable space than a cylindrical containment structure of equal volume.

Mid-Sized Light Water Reactors With Passive Safety Features

The principal U.S. effort to develop mid-sized LWRs with passive safety features is sponsored by EPRI and the U.S. Department of Energy (DOE), with substantial contributions from major U.S. suppliers.[Taylor and Stahlkopf, 1988; Taylor, 1989] (EPRI receives funding from most U.S. utilities and utilities in France, Italy, the Netherlands, Japan, South Korea, and Taiwan.) The passive plant was envisioned as a smaller reactor that would employ primarily passive means--gravity, natural circulation, and stored energy--for its essential safety functions.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The passive LWR design concept was considered potentially attractive to utility investors for several reasons: (1) the fundamental simplicity of the passive safety concept offers an opportunity to effect wholesale simplification (reducing many valves, pumps, pipes, tanks, instruments, etc.), with attendant improvement in construction costs and schedules, and plant operability and maintainability; and (2) by reducing reliance on active components and human intervention, passive features can help accommodate a wide range of upset conditions and internal and external plant threats, such as loss of all electrical power.[Westinghonse, 1989]

It is estimated that, compared to a conventional 600 MWe pressurized LWR, a plant with passive cooling features would offer the following savings in bulk commodities:

  • 60 percent fewer valves;

  • 35 percent fewer large pumps;

  • 75 percent less piping (in the nuclear island, the predominantly nuclear part of the plant);

  • 80 percent less heating, ventilation, and air-conditioning ducting;

  • 80 percent less control cable (nuclear island); and

  • 50 percent less seismic building volume.[Westinghouse, 1989; Taylor and Stahlkopf, 1988]

For a BWR, the following reductions would be achieved:

  • valves by 16 percent;

  • safety-grade pumps and valves by 26 percent;

  • fans by 80 percent; and

  • large pumps by 73 percent.[Taylor, 1989]

The Chairman of the Utility Steering Committee for EPRI's Advanced LWR Program provided the following thoughts on the choice of the 600 MWe size:

This choice was more or less arbitrary. It was arrived at from two directions. The first was that, in discussions with utilities before the ALWR [Advanced LWR] Program began, EPRI concluded that there were a number who felt a smaller size plant in the approximately 600 MWe size range would be better adapted to their system, and would be something more easily accepted, than a plant twice that size. The second reason for the choice was to distance the Passive Plant from the Evolutionary Plant so as to reduce the direct competition between the two.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Enough work has been done to say that 600 MWe is not the limit for gravity removal of decay heat. Both General Electric and Westinghouse have done work, sometimes with Japanese firms, which indicates that plants 900 to 1,000 MWe are feasible. On the other hand, we have not done enough design or experimental work in the United States to say with confidence where a limit rests.

It is true that there is some value established by the size of the reactor vessel or the size of containment which prudently limits the capacity of the first generation of Passive Plants. That is because the power densities are lower and therefore the core size is larger for a given capacity.

I might say that many of us believe this to be an advantage in an overall sense in that we believe one of the problems with present generation plants is that sizes and power densities were pushed too far, too quickly.[Kintner, 1989]

Advanced Passive Pressurized Water Reactor

The advanced passive (AP-600) design was developed by Westinghouse with financial support from DOE and EPRI. Figure 3-4 is a diagram of the AP-600 passive cooling system.[Westinghouse, 1989]

Core Design. The AP-600 has the proven uranium oxide fueled core, with reductions in coolant temperature, flow rates, and core power density to increase design thermal margins.

Steam Generators. The steam generators, of U-tube design, include evolutionary improvements over those in existing plants, including improved tube material to reduce corrosion and upgraded antivibration bars to reduce wear. Lower average coolant temperatures are intended to improve tube integrity. The reactor coolant pumps are mounted in the channel head at the bottom of the steam generator, simplifying the support system, reducing piping and construction, and increasing the space for maintenance.

Fluid Systems. Passive cooling in the AP-600 is achieved with a passive ECCS, which is a combination of two cooling water sources: (1) gravity drain of water from two core makeup tanks and (2) a large refueling water storage tank suspended above the level of the core. Additionally, the ability to inject water from two pressurized accumulator tanks is retained. Core decay heat can also be removed through a passive residual heat exchanger located in the

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-4 AP-600 passive cooling systems. SOURCE: [Westinghouse, 1989]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

refueling water storage tank. This heat exchanger transfers decay heat to the refueling water by natural circulation as shown in Figure 3-4.

Control and Instrumentation. Features of the system include microprocessor-based technology, multiplexed controls for plant data and signal, electrical data links, and fiber-optic data highways. It includes an advanced alarm system, advanced operation display system, and an advanced accident monitoring/safety display system.

Microprocessors and multiplexed data highways permit complete and rapid communication between the central control room and other control and protection cabinets located throughout the plant. Malfunctions anywhere in the plant can be detected and addressed on a real-time basis if plant conditions change from trip setpoints.

Containment. The containment structure is a cylindrical steel shell that, in emergencies, can be cooled by evaporating water, which is gravity-fed from a large tank above the containment structure. This tank holds a three-day water supply and can be refilled externally. Heat is ultimately removed to the atmosphere by a natural air circulation system. Like emergency core cooling, containment cooling requires only automatic valve operations (i.e., no operator action and no pump, diesel, or fan operations) after any major energy release from the maximum LOCA. Concrete shielding is provided external to the steel containment.

Modular Construction. Large-scale studies on the construction of modules are being carried out by Avondale Shipyards and Westinghouse to develop economical assembly techniques in the factory or shipyard.[Taylor, 1989] This construction planning also reflects Japanese experience in fabricating, assembling, and installing large modules in their nuclear plants. It was estimated that these smaller, simpler plants, amenable to factory construction and with the design essentially complete before construction begins, could be built in three to four years following the issuance of a construction permit. This simplified design with shorter construction times and estimated lower capital costs could compensate for the loss of economy of scale credited to larger plants.

Simplified Boiling Water Reactor

The SBWR is a passive design being developed by GE with financial support from DOE and EPRI.[Duncan and McCandless, 1988] Figure 3-5 illustrates this reactor concept.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Core Design. The SBWR's lower power density increases thermal margins in the critical power ratio from 10 percent to more than 30 percent. This indicates that the power at the transition from nucleate to film boiling, relative to the operating power, has increased by 20 percent. The reactor operates at furl power with natural circulation of water so that the recirculation pumps are eliminated, resulting in a simpler reactor vessel, reduced vulnerability to loss of coolant, and reduced maintenance. The larger reactor vessel needed for natural circulation provides the additional benefit of a greater inventory of water above the core at the initiation of any transient conditions.

Fluid Systems. Passive cooling is achieved by locating the suppression pool above the reactor core so that, in an emergency, core cooling is achieved by gravity rather than safety injection pumps. This feature not only eliminates the injection pumps, but also associated valves, piping, and diesel generator power supplies. The suppression pool is a standard feature of current BWRs. It serves as a passive cooling system that reduces the temperature and pressure in the containment building in the event of a severe accident.

During normal operation, an isolation condenser submerged in a pool of water, located above the core and outside the containment, controls reactor pressure passively (automatically) without reducing the fluid volume in the reactor vessel. Isolation condensers for passive reactor pressure control were used in early BWRs and have been reintroduced in this design. These isolation condensers can also be used to remove long-term, postaccident decay heat from the containment. This second passive feature would function in the event of loss of coolant.

Control and Instrumentation. The system includes an advanced control panel design and features an intelligent multiplexing system using fiber optic data transmission and extensive use of standard microprocessor-based control and instrumentation modules. The equipment allows fault-tolerant operation, improved fault detection, and self-diagnostics while the system is in operation.

Containment. The primary containment is a steel-lined reinforced concrete structure with a steel dome located within the reactor building. It is highlighted in black in Figure 3-5. Inside the primary containment is the pressure vessel, gravity-driven cooling system pool, suppression pool, and depressurization valves. The last three provide rapid response in the event of loss of coolant.

Modular Construction. Modularization techniques are proposed to reduce costs and shorten construction schedules to as little as 30 months. These

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×
Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×
  1. Reactor Building

  2. Reactor Building Crane

  3. Refueling Machine

  4. Fuel Handling Machine

  5. Spent Fuel Storage Pool

  6. Spent Fuel Shipping Cask & Pool

  7. Equipment Main Entry Hatch

  8. Isolation Condenser Pool

  9. Isolation Condenser

  10. Reactor

  11. Fine-Motion Control Rod Drives

  12. FMCRD Hydraulic Units

  13. Reactor Pedestal

  14. Under-Vessel Servicing Platform

  15. Lower Drywell

  16. Shutdown Cooling Line

  17. Upper Drywell

  18. Main Steam Lines

  19. Feedwater Lines

  20. Depressurization Valves

  21. Safety Relief Valves

  22. SRV Quenchers

  23. Horizontal Vents

  24. Suppression Pool

  25. Gravity-Driven Cooling Pool

  26. Building HVAC

  27. Control Room

  28. Residual Heat Removal System

    Heat Exchangers

  29. Reactor Component Cooling

    Water System Pump

  30. Reactor Service Water System

    Heat Exchangers

  31. DC Batteries

  32. Plant Stack

  33. FMCRD Electric Panel

  34. Steam Tunnel

  35. Drywell Head

  36. Steam Separator Storage Pool

FIGURE 3-5 The simplified boiling water reactor. SOURCE: [GE Nuclear Energy, 1989]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

techniques will be applied to reinforcing bar assemblies, structural steel assemblies, steel liners for the containment and associated water pools, and selected equipment assemblies, such as isolation condensers, drywell piping, heating, ventilation, and air-conditioning units, and water treatment equipment.

Other Reactor Concepts
CANDU-3 Heavy Water Reactor

HWRs are used in Canada for commercial electric power generation. These reactors are known as CANDU (for Canadian deuterium uranium) reactors. Although DOE operated HWRs for weapons material production for over 30 years, their design is very different from the CANDU design. For example, the current heavy water weapons material production reactor operates at room temperature with no significant pressure, and it has several annuli of fuel within a “universal sleeve homing. ” By contrast, the CANDU is a pressurized reactor, its fuel is within a “pressure tube,” which itself is within a low pressure “calandria tube,” and it operates at a high temperature.

The CANDU-3 is the latest and smallest version of the CANDU pressurized heavy water system developed in Canada.[AECL, Undated; AECL, 1989] Its steam supply system is shown in Figure 3-6. CANDU-3 has a net output of about 450 MWe and complements the established mid-sized CANDU 600 plant. A high level of standardization has been a feature of CANDU reactors. The vendor notes that, in CANDU-3, all key components, such as steam generators, coolant pumps, pressure tubes, and refueling machines, are identical to those in operating CANDU power stations. AECL states that the nuclear safety principles applied to the CANDU-3 reactor ensure that Canadian regulatory requirements are met. These requirements take the form of general criteria against which the developer must establish detailed design requirements.[AECL, Undated]

A letter of intent to submit the CANDU-3 design for standard design certification under 10 CFR Part 52 has been sent to NRC.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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FIGURE 3-6 Steam supply system of CANDU-3. SOURCE: [AECL, Undated]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Core Design. Design of the CANDU-3 reactor core closely follows that of the larger CANDU reactors (of up to 881 MWe). The core design incorporates the standard geometrical arrangements of horizontal fuel-containing pressure tubes in a square lattice and has neutronic characteristics similar to those of current CANDU 600 reactors. There are three features unique to the CANDU designs, including the CANDU-3: (1) the use of natural uranium oxide fuel, (2) the use of heavy water as a moderator and coolant, and (3) on-power fueling.2 The outlet header coolant pressure is about 1,450 psia and the outlet coolant temperature is about 590°F. These operating parameters are somewhat lower than the 2,250 psia and 615°F approximate values of U.S. pressurized LWRs.

The CANDU-3 design has a small positive void coefficient during a large break LOCA, as does the CANDU 600. This coefficient produces a power rise (50 to 100 percent per second) that must be counteracted by one of the two independent shutdown systems.

Steam Generators. CANDU-3 steam generators, like those of the CANDU 600, consist of a vertical U-tube bundle in a cylindrical shell, located above the reactor to ensure natural coolant circulation on loss of power to the primary cooling pumps. As in U.S. PWR systems, the heated coolant (heavy water in CANDU reactors) is contained on the primary side of the steam generator.

Fluid Systems. The ECCS operation includes provisions for both short-term injection from pressurized accumulator tanks and long-term recirculation of a mixture of ordinary and heavy water from the reactor building floor.

Control and Instrumentation. The reactivity control units are the reactor sensor and actuator portions of the reactor regulating and reactor shutdown systems. These systems include reactor power measuring devices, neutron absorbing reactivity control and shutdown devices, and the liquid injection

2  

The CANDU reactor has on-power (also known as “on-line”) refueling, which means that the fuel is changed routinely with the reactor operating at full power. A fueling machine inserts new fuel into the reactor's fuel channels. A fuel transfer system brings new fuel into the reactor building and takes out irradiated fuel. Both the fueling machine and the fuel transfer system are automated and operated from the main control room. Surveillance equipment designed to monitor CANDU refueling operations is used by the International Atomic Energy Agency so that compliance with nuclear safeguards requirements can be verified.[AECL, Undated; AECL, 1989]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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nozzles of the shutdown system. The shutdown system is physically and functionally separate from the regulating system.

All CANDU reactors use digital computers for the control of the reactor regulating system and other process systems, such as the pressurizer and steam boiler levels. In the CANDU-3, however, the two large central computers in the CANDU 600 systems have been replaced by a distributed control system consisting of a number of electronic modules distributed throughout the plant and linked by coaxial-cable data highways. This control system feeds data directly to color graphic operator stations, which form the interface between the operator and the plant. The design also features digital automated startup.

Containment. The CANDU-3 containment consists of a containment envelope of reinforced concrete with a full steel liner. All penetrations that are open to the atmosphere close automatically when an increase in containment pressure or radioactivity level is detected.

Modular Construction. The layout of a CANDU-3 power station permits modular construction because the contents of each building are subdivided into modules on a system and subsystem basis. The interfaces between modules are intended to facilitate site assembly and minimize site construction time. In addition, fuel channels can be factory assembled as can the steel calandria that contain the heavy water moderator. The shield tank, shield tank extension, and deck for the reactivity mechanisms are also amenable to offsite construction.

Safe Integral Reactor

CE has undertaken the design of the SIR jointly with Rolls Royce and Associates Limited, Stone and Webster Engineering Corporation, and the United Kingdom Atomic Energy Authority.[Bradbury et al., 1989] SIR is a PWR in which the reactor core, pressurizer, reactor coolant pumps, and steam generators are contained in a single reactor pressure vessel. The plant can produce a nominal station power output of 640 MWe from one turbine-generator supplied with steam from two identical 320 MWe pressurized LWR modules. Figure 3-7 illustrates the SIR design.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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FIGURE 3-7a The safe integral reactor. SOURCE: [CE, 1989b]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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FIGURE 3-7b The safe integral reactor heat removal systems. SOURCE: [CE, 1989b]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Core Design. The reactor assembly is a completely self-contained PWR within a single vessel. Reactor coolant loop pipes and surge line have been eliminated. The SIR fuel, fuel assembly, in-core and ex-core instrumentation are all patterned after current CE designs for PWRs. By using many small components in parallel within the reactor vessel, primary system connections to the pressure vessel are relatively few and can be kept small; the largest is 2.8 inches in diameter. All pressure vessel penetrations have been kept well above the top of the reactor core.

Steam Generators. Twelve cylindrical steam generator modules are installed in the annular space between the core support barrel and the wall of the reactor pressure vessel. Located above the core, the modules provide the primary circuit natural circulation but are also shielded from the core. Finally, the vendor claims that full power operation can be maintained with one faulty steam generator module isolated.

Control and Instrumentation. This system is based on the System 80+ control and instrumentation design.

Fluid Systems. In the SIR design, there is no primary piping, reducing primary system failures. The normal cooldown process occurs on the secondary side, where subcooled fluid is circulated through the secondary side of the steam generators. For LOCAs, passive decay heat removal systems provide long-term cooling and are configured for a minimum of 72 hours of operation without intervention. The use of soluble boron for reactivity control has been eliminated.

Containment. The containment consists of (1) the reactor vessel compartment, which houses the reactor pressure vessel and support structure; (2) eight cylindrical steel pressure suppression tanks with external fins, each containing a pool of water; and (3) a vent system that connects the reactor vessel compartment to the pressure suppression tanks. The containment structure is filled with inert gas to prevent hydrogen ignition.

The reactor vessel compartment is a steel-lined, reinforced-concrete cylindrical structure capped by a removable steel dome. A vent pipe connects the gaseous space of the compartment to the shop-fabricated, cylindrical, steel pressure suppression tanks. These tanks are housed within a reinforced-concrete structure that has outside air intake and discharge ducts for circulating ambient air.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Modular Construction. The compact and simplified SIR design is suited to modular installation.[Bradbury et al., 1989] With the use of advanced construction techniques, the time from the first concrete pour to fuel loading is estimated to be 30 months.

Modular High-Temperature Gas-Cooled Reactor

In the United States the development of gas-cooled reactors has largely been the result of the efforts of DOE and a group of utilities supporting GA Technologies. The first helium-cooled reactor was a 40 MWe demonstration unit built at Peach Bottom, Pennsylvania, which operated from 1967 to 1974. A type of coated fuel particle was successfully used in this unit. A larger 330 MWe plant was built at Fort St. Vrain, Colorado. This unit was recently shut down because of steam header cracks, a low capacity factor due largely to poor circulator performance, and the resulting poor economics.

In Europe, development work in Germany has been led by Siemens and ABB. France and Great Britain were early pioneers in the use of reactors cooled with carbon dioxide. The German thorium high-temperature prototype reactor (THTR) produced approximately 3 billion kilowatt hours (kWh) of electricity. Some techanical problems occurred during its operation (e.g., high friction of graphite balls). A reevaluation of continued operation of the THTR was made in late 1988. Considerations such as the termination of fuel supply, the inability to assure spent fuel storage, the possibility of additional requirements being imposed prior to obtaining a long-term license, and larger estimated decommissioning costs led the consortium that owns the plant to seek increased government participation or, absent that increase, to shut the reactor down. The THTR was shut down in late 1989.3 [Gas-Cooled Reactor Associates, 1989; Hill, 1989]

The advanced MHTGR concept of GA Technologies is a helium-cooled unit. The important features of the design presented to the Committee are

3  

Germany also operated a 15 MWe pebble-bed high-temperature gas reactor known as the AVR for about two decades. This reactor has been shut down.[Gas-Cooled Reactor Associates, 1989]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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FIGURE 3-8 The advanced modular high-temperature gas-cooled reactor.

SOURCE: [GA, 1989]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

depicted in Figure 3-8.4 This design generates 538 MWe from four nuclear modules and two turbine generators,5 using steam at 2,515 psia and 1,005°F. The high temperature operation of the MHTGR leads to high thermal efficiency.[GA, 1989; Taylor, 1989; Nylan et al., 1988] An additional feature is its potential to provide process heat because of the high coolant exit temperature of 1,268°F. The current development is sponsored by DOE and Gas-Cooled Reactor Associates, with technical support from EPRI. This reactor's nuclear steam supply module is graphite-moderated and helium-cooled. The use of inert helium in contact with graphite core materials leads to low radioactive releases and low radiation exposure to workers in the plant if the helium coolant purity can be satisfactorily maintained. The conceptual design is presently under review by DOE for development as one of two reactor technologies for production of nuclear weapons materials as an eventual successor to the HWRs used at Sawnnah River.

Core Design. The reactor core is a low power density design that consists of an annular array of hexagonal blocks of graphite fuel elements surrounded by a reflector of unfueled graphite blocks. The design is intended to provide efficient heat transfer to the exterior in order to limit the temperature rise of the fuel in the event of a LOCA. The fuel consists of particles of uranium oxycarbide, enriched to about 20 percent in uranium-235, and thorium oxide. The fuel particles or kernels are about 0.8 millimeter in diameter, coated with porous graphite, and covered by successive layers of pyrolytic carbon, silicon carbide, and pyrolytic carbon. The coated particles are bonded together in fuel rods placed within sealed vertical holes in the graphite fuel element blocks.

The graphite fuel element blocks, together with the graphite moderator/reflector, provide a large heat sink in the event of an emergency. Preliminary data from temperature ramp tests of about 50°C per hour indicate essentially no failure of the refractory coating around the fuel particles below1800°C. Coating integrity at elevated temperatures for extended periods requires further evaluation.

4  

The Committee learned in mid-1991 that the MHTGR design has been changed. While the Committee did not have an opportunity to review the new MHTGR study, the Committee understands that the objective was to reduce costs while retaining the postulated safety advantages.[DOE, 1990] Thus, some of the design details listed below may no longer be current (e.g., a given module may produce more power). However, the Committee is not aware of any changes to the fundamental principles underlying the MHTGR concept discussed here.

5  

A possible new design could produce 692 MWe with four somewhat different nuclear modules and four turbine generators.[DOE, 1990]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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LOCA simulation tests were conducted in the late 1980s on a small high-temperature gas-cooled reactor in Germany. The reactor was the experimental 15-MWe AVR (Arbeitsgemeinschaft Versuchs Reaktor). The most significant test demonstrated this reactor's safe response to conditions simulating an accident in which the coolant rapidly escapes from the reactor core and no emergency system is available to restore coolant flow.[Krüger and Cleveland, 1989]

Steam Generator. A single steam generator per reactor module is located in a separate steel vessel. The once-through shell and tube design uses helically wound tubes to carry water in at the bottom and steam out at the top. After passing through superheater sections at the top, the steam is discharged through a nozzle assembly in the upper side wall of the steam generator.[Gas-Cooled Reactor Associates, 1987] Although previous work on the use of a helium turbine in a closed cycle to eliminate the steam generating system was abandoned about ten years ago, developments in high-temperature gas turbines have prompted renewed interest in this concept.

Fluid Systems. If the active cooling system is inoperable in an emergency, decay heat can be dissipated by conduction and radiation to the reactor cavity cooling system in the reactor enclosure. This system circulates atmospheric air by gravity to ultimately remove the decay heat. If the reactor cavity cooling system defaults, passive radiation and conduction transport heat directly to the silo structure and surrounding earth.

Control and Instrumentation. Plant control is based on a fully integrated system in which one operator monitors automatic startup, operation, and shutdown of the two-unit power module. Such a distributed control system would employ the latest technology in computer and communication technology, and system operating procedures would reside in software on local process controllers while overall plant performance was governed by a supervisory computer.[Gas-Cooled Reactor Associates, 1987; EPRI, 1989a]

Containment. The reactor, as presently configured, is located below ground level and does not have a conventional containment. The absence of a containment for the proposed commercial reactor is a major issue, especially given that DOE plans to have a contaiment for the proposed new production

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

reactor version of the MHTGR.6 [Beckjord, 1989] The basic rationale of the designers is that a containment is not needed because of the safety features inherent in the properties of the fuel that were discussed previously. Regarding the possibility of including a containment building for the commercial version, DOE stated to NRC:

Because of its enhanced safety characteristics, the MHTGR has such a high level of safety that no further meaningful improvement in public risk can be obtained at reasonable cost.[Williams et al., 1989]

NRC has not yet made a determination of the acceptability of the proposed MHTGR design without a containment.

Modular Construction. It is claimed that each of the four reactor modules can be factory-fabricated. GA Technologies and Bechtel estimate that

6  

According to DOE, “The primary reason that the MHTGR-NPR [new production reactor] containment system is different from the commercial MHTGR, is to avoid dependence of the development of the NPR on successful completion of the technology program that is necessary to validate assumptions made in the commercial program. The NPR is developing the design and supporting technology in parallel coordinated efforts. These efforts require a decision on the containment system prior to completion of technology efforts that would substantiate the commercial reactor containment approach. The commercial program does not have this constraint.

In addition there are significant differences between the commercial MHTGR and the NPR that justify different design selections to meet requirements. These differences include:

  • Provide the NPR-MHTGR with additional flexibility to accommodate unforeseen future missions.

  • Accommodate the different reactor core which utilizes highly enriched fuel without thorium, includes production materials, and has a different operating cycle.”[Young, 1989]

DOE also stated, “The development of the commercial MHTGR is prepared to be stretched out if there are technology development delays associated with validating the plant design without a low leakage containment structure. Verification of the performance characteristics of high quality fuel is a significant element of the justification for not requiring a low leakage containment structure for the commercial MHTGR.”[Young, 1989]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

construction could be completed in four years from the issuance of a construction permit.[Taylor, 1989]

Process Inherent Ultimate Safety Reactor

The PIUS reactor, a 640 MWe pressurized LWR development, is sponsored by ABB Atom and originated in Sweden. Stated safety features of PIUS are (1) safety ensured by the laws of mechanics and gravity, (2) lack of actively actuated components, (3) lack of required operator action, (4) insensitivity to human errors and malicious intervention, and (5) ability to withstand violent external events.[Bredolt et al., 1988] Figure 3-9 illustrates the PIUS design.[ABB, 1989]

Core Design. PIUS, in the early stages of design, is a passive PWR immersed in a large prestressed concrete pressure vessel filled with cool, borated water at about 1,340 psia.

The reactor is contained in a cylindrical structure that extends from the bottom of the core, near the bottom of the vessel, to the top enclosure. During normal operation this structure separates the circulating hot coolant loop from the cool vessel water by two hydraulic density locks.7 The coolant loop has a low concentration of boron, in contrast to the vessel water. During normal operation, the heat generated in the reactor is carried by the coolant upward to the top of the cylinder and then to a steam generator, where the main coolant pump returns it on a flow path inside the cylindrical structure to a point below the core. The reactor power is controlled by the temperature and the boron content of the reactor circulating loop. There are no control rods in the PIUS 600. If the main coolant pump stops, the water circulates by natural circulation through the density locks, bringing the cool borated water into the core and shutting down the reactor.

The fuel assemblies are standard pressurized reactor fuel elements with low-enriched uranium oxide pellets in fuel rods.

7  

A hydraulic density lock makes use of the principle that water separates naturally into layers that have different densities. The application of that principle in PIUS means that during operation cold borated water sits below the core while lower-boron content hot water in the primary loop passes over the cold water and through the reactor. Loss of circulation in the primary loop results in the cold, highly borated water being drawn into the core through the chimney effect, thus shutting down the reactor.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-9 The process inherent ultimate safety reactor. Main features of the nuclear steam supply system. SOURCE: [ABB, 1989]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Steam Generators. The steam generators, located outside the concrete reactor vessel, use a conventional straight-tube once-through design.

Fluid Systems. Residual core heat can be removed either by the four steam generators or by the pool. Pool heat removal is used for extended shutdowns or in emergency conditions and can be achieved by either passive or active means. In the passive heat removal system, heat exchangers submerged in the pool transfer heat to the secondary side, which is cooled by naturally circulating ambient air drawn from a dry cooling tower. Water temperature can be maintained below 100°C under these conditions even in the case of a large LOCA.

Control and Instrumentation. The predominantly non-safety-grade equipment, based on micro- and mini-computers, is located in the main control room area and is distributed in the plant (decentralized system). The safety-grade parts (e.g., the reactor trip and interlock system with associated measuring systems and control equipment for initiating safety-related actions) are located in two separated compartments at the bottom of the reactor building. All systems are implemented on microcomputers, arranged in redundant trains. Man-machine interactions are based on color video display units with keyboard and tracker balls.

Containment. The containment structure is a large, prestressed-concrete reactor vessel in which the cold borated water, the reactor core/riser assembly, and all key safety systems are located. The key characteristics of this structure are that it (1) contains sufficient borated water to cool the reactor for one week after reactor shutdown, (2) is large enough to store spent fuel for the lifetime of the reactor, (3) provides a high level of protection against saboteurs, (4) contains both steel reinforcing bars and prestressed steel tendons, which together provide a very strong structure, and (5) contains a double internal steel liner to prevent water leakage.

Modular Construction. The construction of PIUS is based on separation of building units, prefabrication of parts of the containment and pressure vessel at the site, use of conventional process systems, limited use of pumps, pipes and cables, and limited scope of equipment located inside the containment. A 36-month construction schedule for the n'th plant is predicted based on BWR construction experience and does not rely on modular construction. A modularization review suggests possibilities for shortening the construction schedule.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×
PRISM Liquid Metal Reactor

Fast reactors normally use liquid sodium as a coolant. They can produce more fissile material than they consume and are often referred to as “breeder” reactors. LMRs have been used to produce electricity in the United States, France, Great Britain, the Soviet Union, and Japan.[Collier and Hewitt, 1987]

In the United States, a small experimental breeder reactor (EBR-I) built by Argonne National Laboratory generated the first electricity from nuclear fission in 1951 and in 1953 confirmed that breeding was possible. In 1955 the second core of this reactor was partially melted during an experiment designed to investigate its prompt positive reactivity feedback coefficient. Fuel rod bowing was determined to be the cause, and subsequent core designs corrected that problem. Subsequently, a second experimental breeder reactor (EBR-II) was built by Argonne National Laboratory in Idaho and began operation in 1963. It has demonstrated the practicality of the LMR design in which the entire primary system is submerged in a pool of sodium. Since the mid-1960s the EBR-II has been a test facility for LMR fuel assemblies and structural material irradiation and safety tests.

In April, 1986 two significant safety tests were conducted at EBR-II. These involved loss of flow without scram from full power and loss of heat sink without scram from full power. These tests successfully demonstrated the safety potential of the integral fast reactor (IFR), a generic reactor technology defined by the use of liquid sodium as coolant and metallic uranium and plutonium as fuel. The reasons for the safe responses illustrated in the EBR-II tests are inherent to the IFR. Specifically, the properties of the metallic fuel and the large thermal inertia of the sodium pool are key to achieving reactor shutdown passively (i.e., without relying on operator intervention, active components such as control rods, pumps, valves, or the use of balance of plant for heat removal) while keeping temperatures low.[Chang, 1989]

While the capability to ride out a loss of flow without scram from full power and a loss of heat sink without scram from full power add markedly to the safety of an LMR, the presence of a positive sodium void coefficient in the present design has been a matter of concern. This has led to the addition of rod stops to control reactivity insertions that may result in sodium

If an advanced LMR is proposed having a significant positive sodium void reactivity worth, careful evaluation will have to be made of the effect of this attribute on the severity of postulated accidents involving reactivity insertion or other events which could lead to sodium boiling. Additional or redundant features may be necessary to remove this concern.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The Committee understands that additional or alternative features to remove this concern are being identified. The positive void coefficient of the currently proposed LMR design [PRISM] results from a design criterion for a shippable reactor vessel that in turn determines the maximum core diameter and the required core height for the specified power module. To eliminate the possibility of the positive void reactivity worth, revisions to the reference PRISM design have been suggested that would result in a larger diameter core with a lower height. Although this arrangement would require a field-fabricated vessel, the elimination of the undesirable positive void coefficient characteristic may be deemed worthy of the loss of a shop-fabricated, rail-shippable reactor vessel.

The more recent fast flux test facility (FFTF) constructed at the Hanford site is a loop-type LMR. It has been used to test full-length oxide fuel assemblies and for limited tests on advanced metallic fuel assemblies. The first commercial LMR built in the United States, Fermi-1, was also a loop-type reactor. It suffered melting of two fuel assemblies in 1966 as a result of a flow-channel blockage. Although Fermi-1 was repaired and restored to operation, it was eventually decommissioned.[Collier and Hewitt, 1987]

The French liquid metal program has constructed three reactors of increasing size culminating in the commercial size SUPER PHENIX plant (SUPER PHENIX was built by a consortium of several European countries). The first two plants, RAPSODIE and PHENIX, performed well and provided valuable experience upon which to build the French program. However, the RAPSODIE experimental reactor was shutdown subsequent to discovery of a tiny leak on the primary sodium circuit, the repair of which was considered too expensive to justify maintaining the reactor in service after 15 years of operation. The 250 MWe demo-plant PHENIX, which started regular operation in 1974, is shut down pending study and evaluation of transient negative reactivity pulses observed in 1989 and 1990 while the reactor was operating at full power. Operation of the SUPER PHENIX 1,200 MWe plant has been curtailed by a sodium leak, discovered in 1987, in an auxiliary vessel for the storage of discharged fuel. This facility is being replaced, and the new one should be ready by the end of 1991.

PRISM is a modular, passively stable, advanced LMR being designed by GE. Its present design uses a new metal alloy fuel being developed concurrently by Argonne National Laboratory as part of the IFR program. The IFR concept includes the first reactor, fuel reprocessing, and fuel fabrication using reprocessed fuel. PRISM is a specific design of generic IFR technology. [Berglund, 1989; Till, 1989] The nuclear steam supply system for a PRISM reactor module is depicted in Figure 3-10.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-10a PRISM nuclear steam supply system and containment. SOURCE: [Griffith, 1988]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-10b Reactor containment. SOURCE: [Griffith, 1990]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Core Design. The PRISM reactor plant is made up of one to three 465 MWe power blocks, each with three 155 MWe reactor modules. (The plant design in ALMR Design and Program Summary and The Liquid Metal Reactor [Berglund, 1989; Till, 1989] uses three power blocks.) The reactor core for each module is in a pool of liquid sodium, which is circulated through the core by four cartridge-type electromagnetic pumps. The pool system consists of a large tank of sodium into which the reactor core, sodium pumps, and two intermediate heat exchangers are placed. The tank is in a guard vessel, which would collect sodium if it were to leak from the pool. This feature assures that the core will remain covered and cooled by sodium.[Berglund, 1989; Till, 1989]

The heat from the reactor module is transferred from the primary sodium coolant loop to a secondary sodium loop through two intermediate heat exchangers. In this way radioactive sodium in the primary loop is isolated from the steam generator. The sodium in the secondary loop enters a single steam generator that produces steam for the turbine generator. The three steam generators for a power block feed steam to a single 465 MWe turbine generator.

The PRISM reference fuel is a uranium-plutonium-zirconium alloy with plutonium concentrations of about 25 percent. As discussed earlier in connection with the safety tests at EBR-II, the properties of metallic fuel are a major contributor to the passive safety features of the PRISM design. Argonne National Laboratory is also developing a pyrometallurgical reprocessing system, in connection with the IFR concept, which could lead to fuel reprocessing and recycling.

Steam Generator. A single-wall helical coil steam generator is believed to provide high reliability (less than one failure per sixty year plant life for a nine unit plant) and economic operation.[Nuclear Power Assembly and ANS, 1990] The steam generator system provides early warning of a tube leak, and an isolation and pressure relief system to limit the sodium-water reaction damage from such a leak, or from multiple tube leaks.

Fluid Systems. The reactor vessel auxiliary cooling system provides emergency core cooling after any incident that impairs the normal emergency heat removal systems. This auxiliary cooling system removes residual heat by radiant heat transfer from the reactor pool to the guard vessel to atmospheric air, which is always circulating upward around the guard vessel. Passive reactor stability is inherent because of a large negative temperature coefficient of reactivity. This combination of passive cooling and passive reactor stability ensures residual heat removal without operator intervention. Thus, if all cooling through the intermediate heat exchangers is lost and the control rods do not automatically shut down the reactor, the negative temperature coefficient of reactivity will bring the reactor to an equilibrium state at a low

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

power level where heat removal from passive systems will maintain the fuel temperature low enough to prevent fuel damage.

Control and Instrumentation. The plant control system, which is not safety grade, provides a high level of automation for normal plant operation and utilizes redundant digital equipment and power supplies to operate nine nuclear steam supplies, three turbine generators, and associated plant equipment from a single control center. Startup, operation, and shutdown of each module are automated. There is a safety grade reactor protection system for each reactor that performs all safety grade functions, including scramming, and is independent of and isolated from the control system.

Containment. The PRISM is located under ground-level. Based on the latest information provided by DOE, the PRISM advanced LMR design includes a lower contaniment vessel and an upper containment dome (see Figure 3-10b). The lower containment is intended to contain reactor pool leaks, while the upper dome is intended to mitigate severe events postulated to cause an expulsion of radionuclides into the region above the reactor. The dome is made of steel that is 1 to 1-1/2 inches thick, and the lower containment consists of 1 inch thick steel.[Griffith, 1990]

Modular Construction. The design includes compact reactor modules sized to enable factory fabrication, economical shipment to both inland and water-side sites, and full-scale prototype testing. Balance of plant modules contain structures and equipment, piping, electrical wiring, and related components.

In Situ Metallurgical Reprocessing of Fuel. Following removal of test fuel pins from the core, an electrorefining process extracts a uranium-plutonium mixture, including fission products producing high dose rates, from the dissolved mixture of fuel, steel, and fission products at temperatures around 550°C. The blanket material is electrorefined in such a way that uranium alone is processed to enrich the product in plutonium. The processed blanket material can then be added to the electrorefined fuel, which is always radioactive.

Actinide Transmutation. Actinides, or the elements in the series beginning with actinium (89) and ending with lawrencium (103), include several very long-lived radioactive alpha emitters and are among the materials of greatest concern in nuclear waste disposal beyond 300 years, depending on the site characteristics and the scenario assumptions under consideration.[Till, 1989;

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

NRC, 1991a] The actinides, in a nuclear reactor or possibly an accelerator designed for the purpose, can be transmuted, with the production of fission energy, to radionuclides that often have a much shorter half-life. An advanced LMR, having no moderator in the core and hence a faster neutron energy spectrum, has much more favorable actinide cross sections than a thermal reactor. An LMR can recycle its own actinides and also actinides from LWR spent fuel, operating as an actinide burner or a breeder, if desired. [Till, 1989; Chang et al., 1987] (A thermal reactor is more limited in the extent to which it can transmute actinides.)

The Committee notes that there exist previous studies of hazards and risks from radioactive waste disposal which have found that, for a given site and a given set of assumptions about repository characteristics and the severity of natural and man-made events, technetium, not the actinides, introduces the greatest risk in the long term.[National Research Council, 1983 and 1984] Since some additional technetium would be the result of recycling the actinides, the net effect would be the production of energy and a proportionate amount of additional technetium, which would still have to be placed in a repository so as to provide long-term safety.

Figure 3-11 is an overview of DOE's proposed actinide recycling process. No such processes for LWR spent fuel recycle have been demonstrated to date, although recycling of plutonium-uranium (mixed oxide) fuel has been demonstrated. Substantial further analysis and research is required to establish (1) whether high-recovery recycling of transuranics and their transmutation can, in fact, benefit waste disposal, and (2) the technical and the economic feasibility of recycling in LMRs actinides recovered from LWR spent fuel.8 [Pigford, 1990] The Committee notes that a study of separations technology and transmutation systems was initiated in 1991 by the DOE through the National Research Council's Board on Radioactive Waste Management.

8  

In late 1990 Professor Thomas Pigford distributed a paper on actinide burning and waste disposal that raised many questions about the technical and economic aspects of recycling actinides in liquid metal reactors.[Pigford, 1990] The Committee has not performed a technical review of that paper but believes that Pigford's analysis supports the need for a careful and objective evaluation of whether the development of transuranic recycle and transmutation, if successful, will actually benefit the geologic repository. Pigford's analysis should be considered carefully by those advocating actinide recycling as a solution to the high-level waste disposal problem. The Committee notes that DOE has provided comments disagreeing with aspects of the Pigford paper.[Young, Undated]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-11 Advanced liquid metal reactor (ALMR) actinide transmutation recycling of light water reactor fuel.

SOURCE: Actinide Recycle, Presentation to National Research Council Committee on Future Nuclear Power Development, Office of Nuclear Energy, U.S. Department of Energy, January 29, 1990

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

EVALUATION OF THE TECHNOLOGIES

The Committee developed the following criteria for comparing the advanced reactor technologies and furnished them to presenters before they briefed the Committee:

  • safety in operation;

  • economy of construction and operation;

  • suitability for future deployment in the U.S. market;

  • fuel cycle and environmental considerations;

  • safeguards for resistance to diversion and sabotage;

  • technology risk and development schedule; and

  • amenability to efficient and predictable licensing.

More detail on the criteria is provided in Appendix B. Vendor estimates related to the criteria are presented in Table 3-3.

The Committee believes that the broad criteria listed above represent the considerations that are (a) most able to be influenced by a choice of technology, and (b) significant to a future determination of whether or not one or more of the advanced reactor technologies is deployed in the United States. For example, the discussion in Chapter 2 has established clearly that the safety and economics of nuclear power substantially affect its acceptance by the public, government, and the private sector.9

The Committee's evaluation was performed by assessing all of the technologies with respect to each broad criterion, starting with safety. (It should be noted that not all subordinate entries in Appendix B were explicitly addressed by the Committee, either because of a lack of specific data or because they were judged to be of lesser importance to the choice of reactor technologies.) The results of the evaluation follow. A summary appears at the end of each section, and the entire evaluation concludes with an overall assessment. The information available with which to perform evaluations is uncertain and often promotional, as should be expected for designs that exist

9  

In the context of studying energy research and development strategies for reducing emissions of greenhouse gases, a National Research Council Committee has suggested that future reactors should be subject to a set of international criteria developed from an international study “on criteria for globally acceptable reactors.”[National Research Council, 1990] Illustrative issues for which criteria would be established include safety, reliability, scale, simplicity and standardization, waste disposal and storage, diversion resistance, cost, and fuel efficiency (i.e., issues similar to those considered in this report).

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

TABLE 3-3 Estimates Provided by Vendors Related to Evaluation Criteria a

Reactor

Core Damage Frequencyj

Availability Goal (percent)

Plant Life (years)

Overnight Capital Costs (1989 $ per rated kWe)g

Levelized Generating Costs (1989 cents per kWh)k

Construction Time (first concrete to on line, in months)n

Projected Date of NRC Certification

Projected Date of Lead Plant Operationb

Large Evolutionary Light Water Reactors

ABWR

<10-6

86

60

940-1,190

3.0-3.3

48

1991e

1996 (in Japan)

APWR-1300

<10-5

90

40

1,350

3.9

54

1995

2000

System 80+ PWR

6×10-7

87

60

<1,143

3.7

54

1992e

Late 1990s (in UK, Korea, or U.S.)

Mid-Sized Light Water Reactors With Passive Safety Features

AP-600 PWR

10-6 to 4×10-6

90

60

1,300

3.9

42

1994e

2000

SBWR

Goal <10-5

90

60

1,200-1,500

3.5-4.1

36

1995e

2000

Other Reactor Concepts

CANDU 3 HWR

Goal <10-6

94

100c

1,713

5.5d

30

1993p

1996 (In Canada)

SIR PWR

6×10-7

87

60

<1,475

<4.3

36

1998

Late 1990s (in UK)

MHTGRm

(note i)

80

Not specified

2,000f

5.2f

44

2002

1998 (Demonstration plant)

PIUS PWR

(note h)

90

60

1,400

4.0

42

Uncertainp

Uncertain

PRISM LMR

To be determined

90

60

1,000-1,375

3.1-5.1

31

2003

1998 (Prototype)

a For comparison with EPRI requirements, see Table 3-2.

b Unless otherwise specified, plant location would be in the United States.

c Modular component life is 40 years.

d Figures provided by AECL were revised to exclude spent fuel management costs, consistent with other vendor estimates.

e The ABWR and System 80+ projections are likely to slip 3 years, and the AP-600 and SBWR projections are likely to slip to 1996, according to NRC staff.[NRC, 1991b]

f These figures are 1988 dollars for an nth of a kind plant.

g Overnight costs exclude time related costs such as interest.

h Vendor has found no credible incidents leading to core degradation.

i PRA implicitly states no fuel-failure sequences down to 10-8 per year. Confirmation of this low probability of fuel failure be sought in a future PRA based on more detailed design information.[Williams, 1989]

j Does not include external initiators such as earthquake and flood.

k Variable costs over several decades (30 to 40 years) are levelized, i.e., converted to equivalent constant annual cost over the same time period. Costs exclude spent fuel management.

m Information provided to the Committee in 1989. A report [DOE, 1990] on a more recent design was subsequently provided. This report, based on the vendor's estimates, indicates improved economics.

n Assumes no licensing delays.

p Late-1990 NRC projections for PIUS indicate 1997 as a possibility; CANDU appears to be extended indefinitely past NRC's earlier projection of late-1996 because its prototype is deferred.[NRC, 1990b; NRC, 1990c]

SOURCES: Vendor presentations to Committee and follow-up communications.The Committee has not evaluated the accuracy of these estimates.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

principally in concept or with non-prototype testbeds. Thus, the Committee concluded that numerical rankings would give a false sense of accuracy. Consequently, the overall assessment represents the Committee's qualitative judgments as a result of considering all the criteria together. The criteria were used primarily in two ways: (1) to provide an outline of issues for the vendors to use in developing their presentations and submissions to the Committee, and (2) to provide a framework for the Committee to discuss the alternative technologies. The Committee concluded that it would not be appropriate to provide weightings for each criterion and then to grade the approaches, add up the scores, and get a selection.

Safety
Discussion

About three-quarters of the nuclear power plants in operation worldwide are situated outside the United States, and this fraction is growing. Another accident anywhere will have major negative consequences for the development of nuclear power worldwide. International cooperation on safety among utilities, suppliers, research organizations, and licensing authorities is therefore necessary.10 The Committee notes that the International Atomic Energy Agengy (IAEA) has established means to monitor the safety performance of nuclear power plants, including the classification of safety significant events (International Nuclear Event Scale).

In the design of future advanced LWRs, vendors are guided by the safety design policy presented in the Requirements Document prepared for EPRI. The safety design policy states that “there will be excellence in safety both to protect the general public and to assure personnel safety and plant investment protection.”[EPRI, 1990] While the safety record of existing nuclear power plants has been very good, more ambitious safety targets have been established for future advanced LWRs. (See Table 3-2 for a summary of EPRI's advanced LWR design requirements.) Safety in the advanced LWR program extends well beyond hardware-oriented lessons learned from existing plants. Attention is focused on areas such as plant simplification, design margins, human factors, and an integrated approach to safety.[EPRI, 1990]

Each reactor designer presented current but only partial design information to the Committee. Where available, probabilistic risk assessments (PRA) were preliminary and did not benefit from detailed system design. Until full

10  

Some believe that next generation nuclear plants will be international efforts subject to international safety standards.[Chung and Hazelrigg, 1989]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

PRAs of actual reactors (including external events) are available and subjected to careful and extensive peer review, there will not be a satisfactory basis to compare the relative safety of the different concepts. Furthermore, the absence of detailed engineering design and the lack of construction and operating experience with the actual reactor concepts make a meaningful, quantitative safety comparison less achievable. In particular, PRA is not a sufficient basis to compare the safety of new concepts with that of proven concepts due to the lack of reliability data of active and passive components sufficiently based on experience. However, if final safety designs of advanced reactors, and especially those with passive safety features, are as indicated to this Committee, an attractive feature of them should be the significant reduction in system complexity and corresponding improvement in operability. While difficult to quantify, the benefit of improvements in the operator's ability to monitor the plant and respond to system degradations may well equal or exceed that of other proposed safety improvements.

The Committee believes that each of the concepts considered can be designed and operated to meet or closely approach the safety objectives currently proposed for future, advanced LWRs, albeit with the considerable uncertainty inherent in risk assessment and in estimates for this extremely low projected level of risk.[Lewis, 1978; NRC, 1990a] If design goals are realized, these plants will be safer than existing reactors. The different advanced reactor designs employ different mixes of active and passive safety features to achieve the safety objectives, and there is, of course, more experience with certain designs than others. The Committee believes that there currently is no single optimal approach to improved safety. There is a distinct advantage to passive containment cooling for preventing containment failure due to slow over-pressurization. However, dependence on passive safety features does not, of itself, ensure greater safety, especially given the potential effects of earthquakes, design errors, inspectability, manufacturing defects, and other subtle failure modes. Consequently, the Committee believes that a prudent design course retains the historical defense-in-depth approach.

In most future reactors, defense-in-depth would be achieved by a multiplicity of safety barriers and features, including a containment structure to mitigate the consequences of core damage accidents. However, one advanced reactor type (the MHTGR), without a containment structure, was proposed. The Committee was not convinced by the presentations or the material supplied to support them that the core damage frequency has been demonstrated to be low enough to make a containment structure unnecessary.11

11  

The Committee notes that, at present, the new production reactor (NPR)-MHTGR program includes a containment. If the MHTGR is selected for the NPR, containment-accident scenario analyses will proceed more rapidly.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×
Summary

The Committee could not make any meaningful quantitative comparison of the relative safety of the various advanced reactor designs. All of the designs are claimed to achieve safety that equals or exceeds the levels specified by EPRI in Table 3-2 (e.g., <10-5/year core damage frequency). If design goals are realized, these plants will be safer than existing reactors. Dependence on passive safety features does not, of itself, ensure greater safety; the historical defense-in-depth approach must be retained. In particular, for the MHTGR, the Committee was not convinced that a containment structure is unnecessary.

Economy
Discussion

Vendor-estimated overnight capital costs and levelized generating costs are shown in Table 3-3 for the reactor technologies that the Committee examined. Most of the estimates for generating costs are based on a 30-year levelized cost analysis, including capital carrying charges, fuel, and operations and maintenance (O&M) (see definitions in Chapter 2). The uncertainties in overnight capital costs and levelized generating costs are quite large because different cost models and assumptions were used for their calculations. Also, U.S. experience with LWRs provides little assurance that construction of the large evolutionary reactors will meet cost and schedule claims.

Vendor estimated overnight capital costs (in dollars per kilowatt electric) and levelized generating costs (in cents per kilowatt hour) for CANDU are higher than those estimated for all LWRs. The higher estimated costs for the CANDU reactor may be partly related to the use of a different cost model than that used by other vendors. Another factor is that CANDUs have been built--the CANDU-3 is quite similar--so AECL has real data to use, unlike some of the other vendors. Additionally, the designs of all the advanced (except possibly for the evolutionary) reactors are still in the stage where cost estimates change. In particular, SIR, MHTGR, PIUS, and PRISM have a very high degree of economic uncertainty. For the different types of evolutionary reactors, levelized generating costs and overnight capital costs are likely to be similar.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

EPRI has independently evaluated some overnight capital costs and O&M costs.[EPRI, 1989b] The estimates are more general than those of the vendors, but they are based on clear definitions. Uncertainties are estimated as -30 to +80 percent. The results in Table 3-4 show that, except for the MHTGR, EPRI's estimates of overnight capital costs are somewhat higher than those of vendors shown in Table 3-3.

The large evolutionary LWRs have higher estimated total construction costs and longer construction times than the mid-sized LWRs with passive safety features, but, as shown in Table 3-3 and Table 3-4, they are estimated to be competitive on a cost per kilowatt electric basis. Estimated construction times under ideal conditions for the mid-sized LWRs with passive safety features ranged from 36 to 42 months (Table 3-3), but there are serious uncertainties about meeting the claimed construction schedules, which in turn could have a major impact on the total funding required to complete the plant. Even though some utilities may prefer to order the larger plants, the perceived larger financial risk may be a deterrent to their deployment.

To reduce construction and operating costs, designers of the advanced mid-sized plants have attempted to simplify their designs, adopt modularized construction, and reduce construction times. However, because there is no experience in building such plants, cost projections for the first plant12 are clearly uncertain. To reduce the economic uncertainties it will be necessary to demonstrate the construction technology and improved operating performance.

Some mid-sized LWRs currently in operation have demonstrated consistently high capacity factors.[IAEA, 1990] Consequently, estimates that assume advanced versions of the same size can also achieve high capacity factors may prove to be correct. (Table 3-3 shows availability projections in the range 80 to 94 percent for all advanced designs and about 90 percent for the mid-sized LWRs with passive safety features. Availability is usually within a few percent of capacity factor. Availability and capacity factor are defined in Chapter 2.) Because the newer heavy water CANDU reactors are a refinement of currently operating reactors, their claimed capacity factors should be attainable. However, capacity factors of the other reactor concepts, SIR, MHTGR, PIUS, and PRISM, have a very high degree of uncertainty.

12  

The descriptive term “first plant” refers to a plant that will be demonstrating new technological features in design, construction, or operations. It is potentially the first commercial operating reactor of this design and, as such, has performance uncertainties in construction and operation. It represents a commercial technology demonstration.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

TABLE 3-4 EPRI-Estimated Overnight Capital and Operations and Maintenance Costs In December 1988 Dollars)

   

Operations and Maintenance Costs

Advanced Reactor Type

Overnight Capital Costs (per rated kWe)

Fixeda ($/kWe-yr)

Incrementalb (cents/kWh)

Large evolutionary light water reactors

$1,300

61.1

0.11

Mid-sized passive light water reactors

$1,475

72.7

0.11

Liquid metal and high-temperature gas-cooled reactors

$1,725

75.5

0.15

a These operating costs are essentially independent of actual capacity factor, number of hours of operation, or amount of kilowatts produced. They include labor charges for plant staff.

b These variable operating costs and consumables are directly proportional to the amount of kilowatts produced. They include chemicals consumed during plant operation.

SOURCE: EPRI. 1989. Technical Assessment Guide, Electricity Supply–1989. EPRI P-6587L, Volume 1: Rev. 6, Special Report, September.

The MHTGRs are estimated to have higher capital costs than the other plants, and they may have higher operating costs, as shown in Table 3-3 and Table 3-4.13 Moreover, if NRC were to mandate a conventional containment, that requirement could adversely affect the economics of this reactor design as well as the technical feasibility of its passive cooling feature.

LMR plants (e.g., PRISM) may be able to compete economically with water reactors if fuel reprocessing (being developed as part of the integral fast reactor program) turns out to be technically and economically feasible, and if the overnight capital costs of these plants are as low as the vendor indicates. (For the IFR, reprocessing would be in situ pyrometallurgical, but for the LMR concept in general, reprocessing options include centralized plants as well as aqueous technology.)[Nuclear Power Assembly and ANS, 1990] EPRI cost estimates (Table 3-4) suggest that these capital costs will be higher.

13  

Recent design changes intended to reduce costs of the MHTGR while retaining its postulated safety advantages imply that the MHTGR economics may be more favorable than reported here. The MHTGR Cost Reduction Study Report [DOE, 1990] states that the modified MHTGR could be cost-competitive with the AP-600. The Committee has not analyzed such projections, nor has EPRI produced a review of them, but notes that they substantiate the large uncertainty in economic projections for advanced reactors. Furthermore, the Committee assumes all reactor designers are working on improvements of the designs and concepts presented to the Committee.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Neither the PRISM design nor the PRISM technology are sufficiently developed to provide a reasonable degree of confidence in cost estimates. Finally, different institutional arrangements may be required for utility involvement in a PRISM plant because of reprocessing, concerns about diversion of sensitive nuclear materials, and lack of utility experience with the technology.

Summary

The economic projections are highly uncertain, first, because past experience suggests higher costs, longer construction times, and lower availabilities than projected and, second, because of different assumptions and levels of maturity among the designs. The EPRI data, which the Committee believed to be more reliable than that of the vendors, indicate that the large evolutionary LWRs are likely to be the least costly to build and operate on a cost per kilowatt electric or kilowatt hour basis, while the high-temperature gas-cooled reactors and LMRs are likely to be the most expensive. EPRI puts the mid-sized LWRs with passive safety features between the two extremes.

Market Suitability
Discussion

None of the reactor concepts the Committee reviewed is likely to be operating in the United States before the year 2000. If the large evolutionary LWRs being built in Japan (and perhaps in Korea) perform well, market potential in the United States will be improved. Large U.S. utilities with several nuclear power plants are likely to be the first customers for such plants if they need large base load electrical generators and if financial risks are acceptable.14

Compared to the large evolutionary reactors, the mid-sized advanced pressurized and simplified boiling water reactors with passive safety features have lower total overnight capital costs (but not lower costs per kilowatt electric), hence less total capital at risk, but no construction and operating experience. The smaller size of these plants might be attractive to a larger number of possible purchasers.

14  

Some Committee members believe that the large evolutionary LWRs will be the next nuclear plants to be ordered in the United States, because of perceived economies of scale and greater confidence by utilities and investors in making modest extensions of proven technology.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The heavy water CANDU reactor has been marketed in Canada and other countries. The main barriers to CANDU's competitiveness in the United States are the uncertainty of its licensing by NRC and the inexperience of U.S. utilities with heavy water technology. On the other hand, the earlier CANDU reactors have a good performance record and could be attractive to certain power producers, particularly if Atomic Energy of Canada, Limited were an investor.[AECL, Undated] It is difficult to weigh all these factors, but the Committee judges that this technology ranks below the advanced mid-sized LWRs in market potential.

The Committee believes there is no near-term U.S. market for the other LWR concepts, SIR and PIUS. While SIR is based on proven light water technology, there are serious uncertainties about the operations, maintenance, economics, and possibly safety of a system configuration that is substantially different from that of current plants. Also, the SIR design appears less complete than the AP-600 or SBWR. The level of testing or prototyping that would be required by NRC is unclear. The PIUS reactor is viewed as a preliminary design with no relevant experience. It is the Committee's view that experience with other LWRs is not relevant to PIUS. While there is no regulatory experience related to PIUS, a conceptual design for this reactor was submitted to NRC for an informal licenseability review. The lack of operational and regulatory experience for both SIR and PIUS is expected to significantly delay their acceptance by utilities, especially if positive experience has been obtained for the evolutionary large reactors or mid-sized LWRs with passive safety features.

The market potential of the MHTGR is very difficult to evaluate. Although gas-cooled reactors have been available for more than 20 years, they have not had commercial success. The strategic advantage of the MHTGR is its high temperature, which permits high temperature process heat applications. However, siting requirements and the extent of a U.S. market for that capability are unclear. The overnight capital cost of the modular design is relatively high on a per kilowatt basis (Table 3-3 and Table 3-4). Further, considerable research and development (R&D) is still required for this advanced reactor, particularly on fuel pellet integrity and on reliable components, and a first plant for demonstration would be required. The issue of whether the design would require a containment building is still not resolved. If no containment building were needed, and the emergency planning zone was reduced to the site boundary, the MHTGR could have significant siting advantages that would make it more competitive. However, based on the Committee's view on containment requirements, and the economics and technology identified above, the market potential for the MHTGR was judged to be low.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

Finally, the LMR might be commercially competitive if uranium fuel shortages limit the use of LWRs. The LMR's safety features and ability to recycle actinides are not considered important positive factors for its early market potential. Any strategy requiring fuel reprocessing introduces significant technical, economic, and non-proliferation police considerations, some of which would complicate licensing.

Summary

The evolutionary LWRs and mid-sized LWRs with passive safety features are judged to have the highest market potential in the United States, while CANDU has the next highest. The other LWR concepts (SIR and PIUS) and the MHTGR are judged to have low U.S. market potential. Finally, the unique properties of the LMR might lead to a U.S. market, but only in the long term.

Fuel Cycle
Discussion

Fuel cycle evaluation encompassed three issues: (1) use of enriched fuel versus use of natural uranium as a fuel, (2) disposal of high-level radioactive waste, and (3) whether fuel reprocessing is needed.15 The environmental implications of the technology derive, in large part, from these fuel cycle issues. Enrichment is important at the front end of the fuel cycle, and the disposal of high-level waste is important at the back end. Reprocessing can influence both the back end (waste disposal) and the front end (need for new uranium fuel). The Committee considered these issues to have roughly equal priority. Again, only enhanced and novel features of advanced reactor designs are discussed.

All LWRs, including SIR and PIUS, have essentially the same fuel cycle and corresponding enviromental implications. Reprocessing of spent fuel to

15  

Reprocessing is not now considered economical in the United States for any reactor technology. Whether it will be so in the future is uncertain. The LMR is the only design that is presently considered for deployment as a breeder, in which event reprocessing would, of course, be necessary. If reprocessing is needed, technical, economic, and non-proliferation issues will have to be resolved.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

recover enriched uranium or plutonium is not currently planned. None of these designs provides a substantially higher burnup than the others.

The CANDU design presented to the Committee uses natural uranium and does not require fuel enrichment; therefore, CANDU does not produce the low-level wastes associated with uranium enrichment. However, it has lower burnup, so the volume of spent fuel rods to be stored will be greater than in the case of the LWRs. In other aspects of fuel cycle management, the heavy water CANDU is comparable to the LWR.

The MHTGR presented to the Committee was designed to use fuel enriched with uranium 235 to about 20 percent versus only a few percent for the LWRs. The fuel pellets provide encapsulation of the waste, which might represent an additional barrier to release of the fission products. However, data to support this have not yet been obtained, nor has a strategy or process for the unique features of MHTGR waste disposal yet been developed. Reprocessing is not currently planned but may become necessary or desirable. At this time, there is no experience with reprocessing of this type fuel, although preliminary development of reprocessing requirements has been investigated. On balance, there does not appear to be a significant fuel cycle advantage or disadvantage to this reactor design.

Finally, the proposed LMR fuel cycle has the potential for substantial economic gains compared to LWR fuels. If a shortage of uranium develops the reactor could breed plutonium.16 However, the feasibility of using this reactor as a breeder in a reprocessing-recycling manner requires policy, technical, and economic development and evaluation. A range of issues needs to be addressed in such a study, including LWR reprocessing as a source of additional fuel and the economics of LWR and MHTGR designs with high conversion ratios.17 Assuming success, it would still be necessary to dispose of high-level waste, although the waste would consist of fission products, most of which, except for technetium, carbon, and some others of little import, have half-lives very much shorter than the actinides.

16  

The LMR breeder could be fueled from stockpiled depleted uranium once it has been started with plutonium or enriched U-235 from some external source. This would remove environmental problems associated with mining uranium and managing the associated mill tailings.

17  

As noted earlier, DOE has initiated a study of separations technology and transmutation systems.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×
Summary

Although there are definite differences in the fuel cycle characteristics of the advanced reactors, fuel cycle considerations did not offer much in the way of discrimination. All LWRs were judged about equal. Compared to the LWRs, CANDUs and MHTGRs had disadvantages at one end of the fuel cycle, but possible advantages at the other. The LMRs offer advantages because of their potential ability to provide a long-term energy supply through a nearly complete use of uranium resources.

Safeguards and Physical Security
Discussion

Safeguards regarding nuclear material in reactors and other facilities must be considered against diversion of fissionable material to nuclear weapons purposes, against sabotage of the power and reprocessing plants leading to a serious accident and release of radioactivity, and against terrorist theft and use of highly radioactive material as a terror weapon.

The problem of diversion is usually considered most serious when the facilities are located in countries that have a motivation for developing nuclear weapons. IAEA has developed an international safeguards regime, including on-site inspections and permanent inspection equipment. The IAEA system is applied to nuclear material at all sites in those non-nuclear weapons states that are party to the Nuclear Non-Proliferation Treaty. It is an obligation of the states to inform IAEA of the relevant sites. This application to all material at all sites is called full-scope safeguards. Many suppliers, including the United States, require such safeguards for exports to any non-weapon states. Other countries, including the nuclear weapon states, have safeguards applied to some, but not all, facilities. It is encouraging that Brazil and Argentina have recently agreed to safeguards on all facilities. The most important constraints for limiting proliferation of nuclear weapons are the political will of non-weapon states to forego weapons-development, the safeguards on nuclear (fissile) materials, and agreements by nations possessing advanced technology not to transfer nuclear weapon-related equipment or knowledge to non-weapon states. However, although such supplier agreements can limit the export of technologies that can be used to develop nuclear weapons, theft of weapons-grade material remains a threat. Accordingly, physical security must be provided for nuclear material, especially when in a form (i.e., enriched uranium or reprocessed plutonium) that is suitable or can readily be made suitable for weapons purposes. Physical security is also vital when nuclear material in storage or transit is susceptible to theft and use for terrorist purposes.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

No country with nuclear weapons, or suspected of having nuclear weapons, has developed these weapons using fissionable material from a civilian power reactor, although civilian power programs have been used as a cover for other activities aimed at developing nuclear weapons. 18 (Dual-use reactors, producing both nuclear-weapons material and electricity, have been used however.) Nevertheless, reactors designed and employed for the production of power remain of concern for proliferation because they can be used for production of weapons grade plutonium. Some power reactors have even been designed to operate with highly enriched uranium or with plutonium as an initially-loaded fuel. Therefore, any fuel cycle must be examined for the possibility of diversion of weapons-grade material, or of material that could be further processed to produce weapons-grade material. In particular, the existence of centrifuge or laser enrichment techniques may make the path to weapons much easier, especially since almost all countries have access to natural uranium. In the future, technologies developed to permit efficient extraction of specific isotopes of plutonium may also facilitate the extraction of that element from spent fuel removed from reactors. This would create additional paths to diversion. In addition, deployment of any new fuel cycle in the United States or any other nuclear weapon state should be examined with a view to avoiding poor precedents in terms of proliferation. Fuel cycles should be designed to minimize diversion opportunities and maximize safeguardability, regardless of the country in which they are implemented.

The once-through fuel cycle where low enrichment fuels are used and the whole fuel rods, together with radioactive fission products, are buried, has the lowest potential for diversion of sensitive nuclear materials. The use of reprocessing19 where plutonium is separated from the radioactive fission products makes the plutonium easier to use, although it must be noted that with normal burnup the presence of Pu240 in the plutonium makes the plutonium much more difficult to use in a reliable bomb. The opportunities for diversion are greater in any concept where on-line fuel loading is possible (i.e., the CANDU). Because it permits an operation where the fuel is removed more easily from the core after only a short time, and without as much Pu240 being built up, the consequences may also be greater.

18  

The Committee recognizes that one or more countries may have used, or may be using, plutonium from power reactors for the production of nuclear weapons. However, the Committee knows of no case where the weapons were initially “developed” using such materials.

19  

Reprocessing could, in theory, be used with any of the reactor concepts under consideration. However, it is required only if the LMR is deployed as a breeder.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

While the high-temperature gas-cooled reactor is envisioned to use higher enriched fuel than the LWRs, the enrichment levels are below weapons-grade. Nevertheless, because much less separations work is needed to convert such material to weapous-grade, its use could increase proliferation concerns.20 The MHTGR's proposed once-through fuel cycle would make the diversion risk not much greater than for the LWRs. In the LMR based on the integral fast reactor concept, reprocessing would take place in a closed system in which fuel containing actinides is produced. This fuel is highly radioactive and requires special handling in such a way that diversion would be difficult. However, the large amount of plutonium in use may require special safeguards, particularly if in situ reprocessing were not used. In particular, LMRs fueled by plutonium would pose a serious safeguards question in countries of proliferation concern. As compared to LWRs, the CANDU reactor poses some additional risks of diversion because of two features: (1) replacing fuel while the reactor is running increases access to fissionable material, especially plutonium, and (2) production and transportation of heavy water provides access to material that is useful in producing weapons-grade material.

Sabotage is always a threat against an industrial facility, posing a risk to the workers and, for some facilities, to the neighboring public. Power reactors pose a hazard because of the large fission product inventory, once the reactor has run for any significant length of time. Sabotage is another way of defeating safety systems. In order to prevent a knowledgeable person, and particularly a knowledgeable group of persons, from causing serious damage to a nuclear power plant by shutting off critical pumps and/or destroying safety systems, appropriate physical security measures must be and are taken. The advanced mid-sized reactors, even with their passive features, do not eliminate the problem of sabotage, but more detailed evaluation of the risks is needed. The other new LWR concepts and MHTGR appear to be even more resistant to a sabotage-induced fission product release. The LMR also has natural barriers to damage from sabotage, but in the reprocessing cycle some significant damage could be done. The concept of “defense-in-depth” that the Committee endorses for other reasons also provides barriers against acts of sabotage, although special design measures are possible that would further reduce the likelihood of successful sabotage without degrading safety.

20  

In raising natural uranium, which contains 0.7 percent U235, to an enriched state containing 93 percent U235, approximately 90 percent of the separative work is expended in reaching an enriched state containing 20 percent U235. Thus, only 10 percent more work is needed to reach the 93 percent U235 state.[American Physical Society, 1978]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×
Summary

The problems of proliferation and physical security posed by the various technologies are different and require continued attention. Safeguards and physical security considerations do not offer much discrimination among the reactor technologies, particularly for deployment in the United States. However, the CANDU (with on-line refueling and heavy water) and the LMR (with reprocessing) will require special attention to safeguards. Reactor designs that have passive safety features, including the high-temperature gas-cooled reactor and LMR, provide additional protection against certain acts of sabotage. However, special attention will be necessary to ensure that the LMR's reprocessing facilities are not vulnerable to sabotage or to theft of plutonium.

Development Risks
Discussion

The large evolutionary LWRs offer the most mature technology and are in various stages of design certification as standardized plants by NRC. The use of mature technology has many advantages:

  • prior experience provides a check on estimates of reliability, maintainability, safety, and economics;

  • construction and operations experience may not be required before NRC certification as a standardized design or before electricity producers would place an order for the design;

  • technically qualified and skilled personnel are currently available;

  • a related infrastructure exists for design, component manufacture, construction, and operation that is directly transferable;

  • to begin operation, required procedures and training would not be drastically different from those used for the most recent LWRs; and

  • almost all current regulations and regulatory experience relate to this type of reactor, so new federal funding is unlikely to be required to complete the certification process, but could accelerate the process.

For mid-sized pressurized and simplified boiling LWRs with passive safety features, successful design certification by NRC may depend in part on the outcome of contractual work recently initiated by DOE. However, regardless of government assistance, some research and much development and design are still required for these reactors. The research and design to demonstrate the passive safety features must be completed before certification. While these reactors are based on many years of LWR experience, they differ from current reactors in construction approach, plant configuration, and safety features. These differences do not appear so great as to require that a first plant be built for NRC certification.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The extent to which additional demonstration will be required by NRC for any new design has not yet been determined. However, in 1991 the Commission stated:

The Commission approves in principle the requirement for prototype testing of new, innovative technology such as the nuclear power plant control room designs intended for design certification, if the testing is required to confirm expected operational performance under normal and abnormal conditions and thus is essential for the [NRC] staff 's safety determination.[Chilk, 1991]

While a prototype in the traditional sense probably will not be required, federal funding will likely be required for the first mid-sized LWR plant with passive safety features. The level of government assistance required to build such a first plant is uncertain but could be significant. The Committee believes that the designs of mid-sized LWRs can be certified by NRC without construction of a prototype plant. However, federal funding is likely to be required to assist in the construction of the first mid-sized LWR plant; such funding would serve to offset some of the factors associated with the innovative features of these designs, such as the risks of not meeting the shortened construction times, the costs of first-of-a-kind engineering, and the uncertainties in the NRC licensing process.

The CANDU-3 reactor is farther along in design than the mid-sized LWRs with passive safety features. However, it has not entered the NRC design certification process. Commission requirements are complex and different from those in Canada so that U.S. certification could be a lengthy process.[Ahearne, 1989] Of particular note is the small positive void coefficient during a LOCA. NRC has always required strong negative void coefficients.

Development risks for the other LWR concepts (SIR and PIUS) are greater than those for the technologies discussed above. Regarding SIR, there is some concern about the reliability of components because access for maintenance is restricted. Extensive design and development are needed, and a full-scale first plant will probably be required before design certification is approved. In addition, numerous technical issues must be resolved to establish the systems' performance during normal operation and the adequacy of the safety features. PIUS incorporates much new technology and has only been demonstrated in laboratory experiments. There is also concern about the stability of the interface between its reactor coolant and the highly borated water in the surrounding pool. At a minimum, a reasonably large experiment that combines neutronics with thermal hydraulics would be required to alleviate this concern, but the Committee believes a full-scale first plant will probably be necessary.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The MHTGR needs an extensive R&D program to achieve commercial readiness in the early part of the next century. The construction and operation of a first plant would likely be required before design certification. Although there is worldwide experience with gas-cooled reactors, most of these reactors are sufficiently different from the MHTGR that much of this experience is not relevant to the technical uncertainties relating to the advanced reactor type.21 Experience with the U.S. Fort St. Vrain reactor and the German THTR underscores the need to complete development and build a first plant to identify potential problems in a full-scale plant. The advanced gas-cooled reactor is claimed to have a unique safety feature in its encapsulated fuel particles. However, additional R&D would be needed to confirm that fission product containment in mass-produced, core quantity batches is achieved at severe accident temperatures of 1,800° to 2,000°C for extended periods with extremely high reliability. Oak Ridge National Laboratory estimates that data to confirm fuel performance will not be available before 1994.[Homan, 1989] The Committee also recognizes that the particle fuel concept of the MHTGR may lead to significant focus of regulatory safety inspection on the common fuel manufacturing facility, because of the required stringent quality of the approximately nine billion particles comprising each core and the maintenance of this regime throughout the operating lifetime of the manufacturing facility. Means to achieve assurance will have to be developed. The Committee believes that reliance on the defense-in-depth concept must be retained, and accurate evaluation of an advanced reactor's safety profile will require evaluation of a detailed design. Studies of accident scenarios should be continued, including the effect of air ingress accidents on the structural support of the core to assure that the core configuration does not change. Finally, the MHTGR does offer the unique capability of producing high-temperature process heat, but to achieve this potential, an extensive development program involving major components must be successfully completed.

The LMR already has an operating test bed reactor (EBR-II), an operating irradiation test facility (FFTF), and a well-framed program to develop LMR technology and demonstrate the integral fast reactor concept. Results to date are promising, and a modular plant design is being developed. This program is backed by a long history of LMR R&D in the United States. However, much R&D is still required. A federally funded program, including one or more first plants, will be required before any LMR concept would be

21  

If the MHTGR is selected for the new production reactor, substantial development funding for the military production version would also benefit the civilian version. The vendor association estimates that such benefits could amount to about $0.4 to $0.7 billion (i.e., a reduction from about $1 billion to $0.3 - $0.6 billion).[DOE, 1990]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

accepted by U.S. utilities. For example, there is lingering concern about the use of sodium because of the possibility of sodium-water reactions and potential fire hazards, although relevant experience with such reactors to date has been positive on these points. There also is some safety concern about the large positive sodium void coefficient in some core designs, although the overall temperature and power coefficient are negative. The opacity of sodium makes the assurance of satisfactory in-vessel inspections and operations more difficult. An accident that produces significant core-wide boiling is very unlikely. In addition, containment is designed to withstand such an accident, including fuel melting.[Nuclear Power Assembly and ANS, 1990] Finally, in situ fuel reprocessing22 must be demonstrated, and concerns about proliferation must be allayed. The economics of this technology, including costs of reprocessing facilities, can be demonstrated only after a first plant is built and operating.

Summary

The large evolutionary LWRs are judged to have the least development risk. The CANDU-3 reactor is farther along in design than the mid-sized LWRs with passive safety features. However, it has not entered NRC 's design certification process. For these designs it is probable that a first plant will not be required for certification. However, the Committee believes that, while a prototype in the traditional sense will not be required, federal funding will likely be required for the first mid-sized LWR plant with passive safety features to be ordered. The remaining reactor technologies have significant development risk, and all will require a federally supported first plant.

Licensing
Discussion

The large evolutionary LWRs are furthest along in the design certification process. They clearly should be most amenable to efficient and predictable licensing and will very likely be the first to be certified. For the mid-sized LWRs with passive safety features, EPRI is working closely with the industry to help move the licensing process forward. These reactors are likely to be the next type certified.

22  

It is possible that centralized reprocessing may be selected instead of in situ reprocessing.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The CANDU reactor can probably be licensed in this century, although it is probably farther behind in the process than the mid-sized LWRs with passive safety features. Moreover, obtaining certification for the CANDU could require substantial additional work on the part of the developer because of great differences in Canadian and U.S. regulatory systems.[Ahearne, 1989]

The SIR and PIUS reactors are still farther behind in the licensing process, and much R&D would have to be done before they could apply for certification. However, these reactors appear to be certifiable eventually, although a first plant will probably be needed. With adequate funding to complete the development program, a demonstration plant for the MHTGR could be licensed slightly after the turn of the century, with certification following demonstration of successful operation. The LMR based on the integral fast reactor concept is still in a very early stage, with much new technology to be evaluated. Reprocessing and recycling will raise significant licensing issues. From the viewpoint of commercial licensing, it is far behind the evolutionary and mid-sized LWRs with passive safety features in having a commercial design available for review.

Summary

It would appear that the large evolutionary LWRs could obtain a NRC design certification as soon as the early to mid-1990s, and the mid-sized LWRs with passive safety features perhaps a little later, followed by CANDU. First plants will probably be required for the other reactor concepts, whose design certification would not be forthcoming until perhaps a decade or more later. The alternative R&D programs presented in Chapter 4 reflect these judgments.

Overall Assessment

The Committee's overall assessment of these technologies is that the large evolutionary LWRs and the mid-sized LWRs with passive safety features rank highest relative to the evaluation criteria. The evolutionary reactors could be ready for deployment by 2000, and the mid-sized could be ready for initial plant construction soon after 2000. The mature evolutionary designs would be available if significant new nuclear generating capacity should be needed before the mid-sized LWRs are ready. Both types of LWRs take advantage of the extensive experience with current reactors, yet they promise improvements in the most troublesome aspects of that experience (e.g., cost, schedule, and licensing). Determinants of the choice among these systems would be perceived financial risk and associated financial arrangements, capacity requirements, and availability of certified, standardized designs.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

The heavy water reactor is also a mature design, and Canadian entry into the U.S. marketplace would give added insurance of adequate nuclear capacity if it is needed in the future. But the CANDU does not offer advantages sufficient to justify U.S. government assistance to initiate and conduct its licensing review.

The other LWR concepts (SIR and PIUS), the MHTGR, and the advanced LMR are believed to be considerably less mature and hence not likely to be deployed for commercial use in the United States until perhaps 2010 to 2025 or later, assuming their development proceeds. SIR and PIUS primarily offer safety benefits. The advanced gas-cooled reactor offers safety benefits and the potential of producing process heat. The advanced LMRs are also judged to offer benefits in their safety and in their ability to breed fuel should uranium resources become scarce. Their potential to alleviate some of the waste disposal problem for LWR fuel through actinide recycling is in such a preliminary stage that this feature is not considered justification for advancing the advanced LMR development program nor delaying waste repository schedules. The Committee judges that the MHTGR process heat capability is of little strategic significance compared with the LMR's potential for breeding. Based on information available at the time of the Committee 's review, the Committee did not judge the safety benefits among the reactors discussed in this paragraph to be significantly different, and thus safety is not a discriminant. The development required for commercialization of any of these concepts is substantial.

The Committee's evaluations and overall assessment are summarized in Figure 3-12.

The Committee's major conclusions regarding the advanced reactor technologies flow from the above assessment. These conclusions are as follows:

  1. Safety and cost are the most important characteristics for future nuclear power plants.

  2. LWRs of the large evolutionary and the mid-sized advanced designs offer the best potential for competitive costs (in that order).

  3. Safety benefits among all reactor types appear to be about equal at this stage in the design process. Safety must be achieved by attention to all failure modes and levels of design by a multiplicity of safety barriers and features. Consequently, in the absence of detailed engineering design and because of the lack of construction and operating experience with the actual concepts, vendor claims of safety superiority among conceptual designs cannot be substantiated.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

FIGURE 3-12 Assessment of advanced reactor technologies.

This table is an attempt to summarize the Committee's qualitative rankings of selected reactor types against each other, without reference either to an absolute standard or to the performance of any other energy resource options. This evaluation was based on the Committee's professional judgment.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×
  1. LWRs can be deployed to meet electricity production needs for the first quarter of the next century:23

  1. The evolutionary LWRs are further developed and, because of international projects, are most complete in design. They are likely to be the first plants certified by NRC. They are expected to be the first of the advanced reactors available for commercial use and could operate in the 2000 to 2005 time frame. Compared to current reactors, significant improvements in safety appear likely. Compared to recently completed high-cost reactors, significant improvements also appear possible in cost if institutional barriers are resolved. While little or no federal funding is deemed necessary to complete the process, such funding could accelerate the process.

  2. Because of the large size and capital investment of evolutionary reactors, utilities that might order nuclear plants may be reluctant to do so. If nuclear power plants are to be available to a broader range of potential U.S. generators, the development of the mid-sized plants with passive safety features is important. These reactors are progressing in their designs, through DOE and industry funding, toward certification in the 1995 to 2000 time frame. The Committee believes such funding will be necessary to complete the process. While a prototype in the traditional sense will not be required, federal funding will likely be required for the first mid-sized LWR with passive safety features to be ordered.

  3. Government incentives, in the form of shared funding or financial guarantees, would likely accelerate the next order for a light water plant. The Committee has not addressed what type of government assistance should be provided nor whether the first advanced light water plant should be a large evolutionary LWR or a mid-sized passive LWR.

  1. The CANDU-3 reactor is relatively advanced in design but represents technology that has not been licensed in the United States. The Committee did not find compelling reasons for federal funding to the vendor to support the licensing.

  2. SIR and PIUS, while offering potentially attractive safety features, are unlikely to be ready for commercial use until after 2010. This alone may limit their market potential. Funding priority for research on these reactor systems is considered by the Committee to be low.

  3. MHTGRs also offer potential safety features and possible process heat applications that could be attractive in the market place. However, based on the extensive experience base with light water technology in the United States, the lack of success with commercial use of gas technology, the likely

    23  

    While this may lock the U.S. into LWR technology for the next 20+ years, the reasons for which are summarized in the following paragraphs, it does not discourage research and development of competitive technologies which may be needed later, as described in Chapter 4.

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

higher costs of this technology compared with the alternatives, and the substantial development costs that are still required before certification,24 the Committee concluded that the MHTGR had a low market potential. The Committee considered the possibility that the MHTGR might be selected as the new tritium production reactor for defense purposes and noted the vendor association's estimated reduction in development costs for a commercial version of the MHTGR. However, the Committee concluded, for the reasons summarized above, that the commercial MHTGR should be given low priority for federal funding.

  1. The LMR technology also provides enhanced safety features, but its uniqueness lies in the potential for extending fuel resources through breeding. While the market potential is low in the near term (before the second quarter of the next century), it could be an important long-term technology, especially if it can be demonstrated to be economic. The Committee believes that the LMR should have the highest priority for long-term nuclear technology development.

  2. The problems of proliferation and physical security posed by the various technologies are different and require continued attention. Special attention will need to be paid to the LMR.

The above conclusions formed the basis for the formulation of alternative U.S. R&D programs in Chapter 4.

24  

The Gas Cooled Reactor Associates estimates that, if the MHTGR is selected as the new tritium production reactor, development costs for a commercial MHTGR could be reduced from about $1 billion to $0.3 - $0.6 billion.[DOE, 1990]

Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

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Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
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Suggested Citation:"3 Assessment of Advanced Nuclear Reactor Technologies." National Research Council. 1992. Nuclear Power: Technical and Institutional Options for the Future. Washington, DC: The National Academies Press. doi: 10.17226/1601.
×

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NRC. 1991a. Transcript of 30th Meeting of the Advisory Committee on Nuclear Waste Bethesda, Maryland. April 24, 1991.

NRC. 1991b. Transcript of meeting entitled Briefing on Progress of Design Certification Review and Implementation. Accompanying viewgraphs. June 12, 1991.

Nuclear Power Assembly and ANS. 1990. PRISM, the Plant Design Concept for the U.S. Advanced Liquid Metal Reactor Program. Paper given at the Nuclear Power Assembly in Washington, D.C. in May and ANS Conference in Nashville, Tennessee in June.

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×

Pigford, T.H. 1990. Department of Nuclear Engineering University of California Berkeley. Letter to Archie L. Wood. National Research Council. November 19, 1990. Transmitting copy of paper entitled “Actinide Burning and Waste Disposal”. An Invited Review for the MIT International Conference on the Next Generation of Nuclear Power Technology. UCB-NE-4176. October 5, 1990.

Taylor, J. J. and K.E. Stahlkopf. 1988. (no title). Nuclear Engineering Design. 109:19(September-October).

Taylor, J. J. 1989. Improved and Safer Nuclear Power. Science. 244:318-325. April 21, 1989.

Till, C.E. 1989. The Liquid Metal Reactor. Overview of the Integral Fast Reactor Rationale and Basis for Its Development. Presentation to National Academy of Sciences Committee on Future Nuclear Power Development. Argonne National Laboratory. August 21-25, 1989.

Westinghouse Electric Corporation. 1989. Assessing the Merits of AP600 Advanced Reactor Technology for U.S. Electric Power Needs. Report to the Committee on Future U.S. Nuclear Power Development. (Energy Systems Business Unit, AP600 Program Office). August.

Williams, P.M., T.L. King, and J.N. Wilson. 1989. Draft Preapplication Safety Evaluation Report for the Modular High-Temperature Gas-Cooled Reactor. NUREG-1338. Division of Regulatory Applications. U.S. Nuclear Regulatory Commission. Washington, D.C. 20555. March.

Wolfe, B. and D.R. Wilkens. 1988. Improvements in Boiling Water Reactor Designs and Safety. Presented at the American Nuclear Society Topical Meeting. Seattle, WA. May 1-5, 1988.

Young, W. H., Assistant Secretary for Nuclear Energy, DOE. 1989. Letter to Eric Beckjord. Director. Office of Nuclear Regulatory Research. U.S. Nuclear Regulatory Commission. November 28, 1989. Enclosure entitled Summary - Containment Study for Modular High Temperature Gas-Cooled Reactor (MHTGR).

Young, W. H. Assistant Secretary for Nuclear Energy, DOE. Undated. Letter to Professor Thomas H. Pigford. (Letter was undated, but received by the Committee on March 5, 1991).

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The construction of nuclear power plants in the United States is stopping, as regulators, reactor manufacturers, and operators sort out a host of technical and institutional problems.

This volume summarizes the status of nuclear power, analyzes the obstacles to resumption of construction of nuclear plants, and describes and evaluates the technological alternatives for safer, more economical reactors. Topics covered include:

  • Institutional issues—including regulatory practices at the federal and state levels, the growing trends toward greater competition in the generation of electricity, and nuclear and nonnuclear generation options.
  • Critical evaluation of advanced reactors—covering attributes such as cost, construction time, safety, development status, and fuel cycles.

Finally, three alternative federal research and development programs are presented.

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