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.



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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE FIGURE 3-1 Advanced boiling water reactor, pressure vessel and core. SOURCE: GE Nuclear Energy

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.]

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE FIGURE 3-3 Elevation view of System 80+ containment. SOURCE: [CE, 1989a]

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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: Safety and cost are the most important characteristics for future nuclear power plants. LWRs of the large evolutionary and the mid-sized advanced designs offer the best potential for competitive costs (in that order). 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE LWRs can be deployed to meet electricity production needs for the first quarter of the next century:23 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. 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. 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. 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. 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. 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.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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. 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. 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]

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE REFERENCES ABB. 1989. PIUS, Presentation Material for Committee on Future Nuclear Power Development, ABB Atom. July 26, 1989. AECL. Undated. CANDU 300 Technical Outline. Revision Seven. Document PPS-74-01010-003. Received 1989. AECL Technologies. 1989. Presentation by AECL Technologies to the National Academy of Sciences ' Committee on Future Nuclear Power Development Reactor Technologies Workshop and Accompanying Documents. Irvine, CA. August 23, 1989. Ahearne, J. F. 1989. A Comparison Between Regulation of Nuclear Power in Canada and the United States. Progress in Nuclear Energy. 22: 3. 1988. Received May 16, 1989. American Physical Society. 1978. Reviews of Modern Physics. Report to the APS by the Study Group on Nuclear Fuel Cycles and Waste Management. 50: 1. Part II. January. Beckjord, E. 1989. Nuclear Regulatory Commission's Director of Research. Letter to Mary Ann Novak, Acting Assistant Secretary for Nuclear Energy DOE. February 28, 1989. Berglund, R.C. 1989. ALMR Design and Program Summary. General Electric (Presentation to Committee on Future Nuclear Power Development. August). Bradbury, R., J. Longo, R. Strong, and M. Hayns. 1989. The Design Goals and Significant Features of the Safe Integral Reactor Presented at the ANS 1989 Annual Meeting. Atlanta, Georgia. June 4-8, 1989. Bredolt, Ulf et al. 1988. PIUS-The Next Generation Water Reactor. American Nuclear Society Conference: Safety of Next Generation Power Reactors. Seattle, Washington. May. CE. 1989a. System 80+™ Advanced Light Water Reactor Technology Assessment, A Report to the Committee on Future U.S. Nuclear Power Development CE. 1989b. Simplified Passive Advanced Light Water Reactor Plant Program. Safe Integral Reactor (SIR) Technology Assessment. A Report to the Committee on Future U.S. Nuclear Power Development

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE Chang, Y.I., M.J. Lineberry, L. Burris, and L.C. Waters. 1987. Nucl. Eng. Int. 32: 23 (November). Chang, Y.I. 1989. Integral Fast Reactor Technology. Presentation to National Academy of Sciences Committee on Future Nuclear Power Development. Argonne National Laboratory. August 21-25, 1989 Chilk, S. J., Secretary, U.S. Nuclear Regulatory Commission. 1991. Memorandum for James M. Taylor, Executive Director for Operations Subject: SECY-90-377. Requirements for Design Certification Under 10 CFR Part 52. February 15, 1991. Chung, K., and G.A. Hazelrigg. 1989. Nuclear Power Technology: A Mandate for Change. Nuclear Technology. 88(November). Collier, T.G., and G.F. Hewitt. 1987. Introduction to Nuclear Power. Hemisphere Pub. Corp. DOE. 1990. HTGR. MHTGR Cost Reduction Study Report. DOE-HTGR-88512. Issued by Gas Cooled Reactor Associates. October. Duncan, J. D. and R. J. McCandless. 1988. Safety of Next Generation Power Reactors. A Paper Presented at the ANS Topical Meeting. Seattle, WA. May 1-5, 1988. EPRI. 1990. Advanced Light Water Reactor Utility Requirements Document. ALWR Policy and Summary of Top-Tier Requirements. 1(March). EPRI. 1989a. Utility Industry Evaluation of the Modular High Temperature Gas-Cooled Reactor. August. EPRI. 1989b. Technical Assessment Guide: Electricity Supply--1989. EPRI P-6587-L. 1:(September). Rev. 6. Special Report. GA. 1989. Modular High Temperature Gas-Cooled Reactor Technology and Applications A Presentation to the National Academy of Sciences' Committee on Future Nuclear Power Development. August 23 and 24, 1989. Gas-Cooled Reactor Associates. 1987. A Utility/User Summary Assessment of the Modular High Temperature Gas-Cooled Reactor Conceptual Design. Nov. GCRA Report 87-011. Revision 1. Gas-Cooled Reactor Associates. 1989. Eleventh International HTGR Conference Overview. News Letter. Summer.

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NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE GE Nuclear Energy. 1989. GE Advanced LWR Technology, Presentation to National Academy of Sciences Committee on Future Nuclear Power Development, August 22, 1989. Griffith, J. D. 1988. Advanced Reactor Program Presentation to the National Research Council U. S. Department of Energy Nuclear Research and Development Program December. Griffith, J., Associate Deputy Assistant Secretary for Reactor Systems Development and Technology, Office of Nuclear Energy, DOE. 1990. Letter to Norm Haller, National Research Council. December 14, 1990 Containing drawings and extracts from Section G.4.1 of Appendix G. Amendment 13 to the PRISM (ALMR) Preliminary Safety Information Document. May. Hill, Phil. 1989. Germany Shuts Down Two New Nukes. Environmental Action. November/December. p. 17. Hirata, K., M. Negishi, C. Matsumoto et al. 1989. Advanced Pressurized Water Reactor Plant. Nuclear Europe. 11-12. Homan, F. 1989. Readiness of MHTGR Technology for Commercial Development. Presentation to the National Research Council. National Academy of Sciences. Committee on Future Nuclear Power Development. Irvine, CA. August 24, 1989. International Atomic Energy Agency. 1990. Nuclear Power Reactors in the World. Reference Data Series No. 2. April Edition. Kintner, E. E. 1989. Letter to Archie L. Wood. November 9, 1989. Krüger, K. (AVR-FRG) and Cleveland, J. (ORNL). 1989. Loss-of-Coolant Accident Experiment at the AVR Gas-Cooled Reactor Thermal-Hydraulic Aspects of Passive Safety and New Generation Reactors-III Winter Meeting of the American Nuclear Society. Transactions of the American Nuclear Society. 60: 735-736. November. Lewis, H.W. 1978. Chairman. Risk Assessment Review Group Report to the U.S. Nuclear Regulatory Commission. NUREG/CR-0400. September. McCutchan, D., T. van de Venne, and J. Cobian. 1989. Improved Availability and Operation in the Advanced Pressurized Water Reactor (APWR). Presented at the International Conference on Availability Improvements in Nuclear Power Plants. Madrid, Spain. April.

OCR for page 91
NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE National Research Council. 1990. Confronting Climate Change. Strategies for Energy Research and Development. August. National Research Council. 1984. A Review of the Swedish KBS-3 Plan for Final Storage of Spent Nuclear Fuel. National Academy Press. Washington, D.C. National Research Council. 1983. A Study of the Isolation System for Geologic Disposal of Radioactive Wastes. National Academy Press. Washington, D.C. NRC. 1990a. Advisory Committee on Reactor Safeguards. Letter to Chairman Carr. Subject: Review of NUREG-1150, “Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants”. November 15, 1990. NRC. 1990b. Memorandum for Chairman Carr and Commissioners Rogers, Curtiss, and Remick from James M. Taylor. Executive Director for Operations. Subject: Response to Items 1 and 2 of the SRM of 10/02/90, “Support for Reviews of CANDU 3 and PIUS Designs.” November 23, 1990. NRC. 1990c. Memorandum for Chairman Carr and Commissioners Rogers, Curtiss, and Remick from James M. Taylor. Executive Director for Operations. Subject: Response to Items 1 and 2 of the SRM of 10/02/90, “Support for Reviews of CANDU 3 and PIUS Designs.” December 17, 1990. Enclosure: Memorandum to A.T. Gody from J. G. Giitter. Subject: Summary of Working Meeting with Atomic Energy of Canada Limited Technologies. November 28, 1990 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. Nylan, A. et al. 1988. In Proc. of the Intersociety Energy Conversion Engineering Conference Denver, CO. August (American Society Mech. Engr. New York). pp. 483-488.

OCR for page 91
NUCLEAR POWER: TECHNICAL AND INSTITUTIONAL OPTIONS FOR THE FUTURE 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).