Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████

NEW REACTOR CONCEPTS FOR NEW GENERATION OF PLANTS IN THE USA: AN OVERVIEW

JASMINA VUJIĆ, EHUD GREENSPAN and MIODRAG MILOŠEVIĆ*

Department of , University of California, Berkeley, CA, USA [email protected], [email protected] *VINCA Institute of Nuclear Sciences, Belgrade, Serbian and Montenegro [email protected]

ABSTRACT

With the growing demands for more reliable energy sources, there is an international interest in the development of new nuclear energy systems to be deployed between 2010 and 2030, that will improve safety and reliability, decrease proliferation risks, improve management and lower cost of nuclear energy production. Six nuclear energy systems were selected as candidates for this Generation IV initiative. In this paper we will explore each of these concepts, as well as several of more advanced concepts.

Key words: Generation IV, nuclear energy systems, non-proliferation

1. INTRODUCTION

Just a few years ago, many analysts were projecting the decline of nuclear power. Reality has proven these projections to be wrong. The reasons for revival of nuclear energy are very clear:

• nuclear plants have performed exceedingly well. As a group, U.S. nuclear utilities have improved the availability of their plants from about 70 percent in 1990 to close to 90 percent today and are producing at about 2 cents per kilowatt-hour (about the same as the most efficient natural gas plants); • consolidation of the nuclear utility industry is leading to the formation of large nuclear utilities with tremendous efficiencies and expertise in operations, maintenance, and training and who have a long-term interest in nuclear power; and • the NRC has reformed its operation and has, with the success of its license renewal process, proven itself to be a fair and consistent regulator with which industry can work to continue operating new plants and potentially build new ones.

Uranium resources would be depleted within a few decades without the deployment of fast breeder reactors in symbiosis with the existing and new reactor types. Thus, for a transition to a nuclear driven hydrogen economy to be sustainable, the nuclear side of the equation must include not only the tie to hydrogen generation but also to the breeding of fissile from fertile

Opportunities for new construction include:

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 495 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ • Evolutionary Light Water Reactors; Advanced designs already certified by NRC, System 80+, ABWR. • Generation III+; Designs that can be certified and built in this decade, Developed under the DOE ″Near-Term Deployment″ effort, AP-1000, ESBWR, PBMR. • Generation IV; Advanced, integrated nuclear energy systems, Developed under the DOE ″Generation IV Roadmap″ effort, With goals, ; safety and reliability and economics.

The current LWRs (identified by the DOE as Generation II reactors) cope and interfere with accident sequences through active means to assure that the consequences of the accident remain within specified acceptable limits.

Advanced reactors now being considered for deployment (or Generation III and III+), like AP600/AP1000, adopt the same philosophy, but accomplish it with passive means to the maximum extent possible.

Concerns over energy resource availability, climate change, air quality, and energy security suggest an important role for nuclear power in future energy supplies. While the current Generation II and III designs provide an economically, technically, and publicly acceptable electricity supply in many markets, further advances in nuclear energy system design can broaden the opportunities for the use of nuclear energy. To explore these opportunities, the U.S. Department of Energy's Office of Nuclear Energy, Science and Technology has engaged governments, industry, and the research community world-wide in a wide-ranging discussion on the development of next- generation nuclear energy systems known as "Generation IV". This has resulted in the formation of the Generation-IV International Forum (GIF), a group whose member countries are interested in jointly defining the future of nuclear energy research and development. In short, "Generation IV" refers to the development and demonstration of one or more Generation IV nuclear energy systems that offer advantages in the areas of economics, safety and reliability, sustainability, and could be deployed commercially by 2030. The Generation IV reactors are supposed to demonstrate enhanced safety with respect to the passive designs [1-4].

2. TECHNOLOGY GOALS FOR GENERATION IV NUCLEAR ENERGY SYSTEM

The guiding principles are: • technology goals for Generation IV systems must be challenging and stimulate the development of innovative systems; • Generation IV systems must be responsive to energy needs worldwide; and • Generation IV concepts must define complete nuclear energy systems, not simply reactor technologies.

Caveats to the goals:

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 496 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ • the goals will guide the development of new nuclear energy systems on a broad front. While desirable, not all goals may be met by any single nuclear energy system; • the goals are not overly specific because the social, regulatory, economic, and technological conditions of 2030 and beyond are difficult to predict; and • the goals must not be construed as regulatory requirements.

The goals include three goal areas:

• Sustainability Resource input SU-1 Waste outputs SU-2 Nonproliferation SU-3

• Safety & Reliability Excellence SR-1 Core damage SR-2 Emergency response SR-3

• Economics Life cycle cost EC-1 Risk to capital EC-2

SU-1: Generation IV nuclear energy systems including fuel cycles will provide sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production. SU-2: Generation IV systems will minimize and manage their nuclear waste and notably reduce the long term stewardship burden in the future, thereby improving protection for public health and the environment. SU-3: Generation IV nuclear energy systems including fuel cycles will preserve the proliferation resistance.

SR-1: Generation IV nuclear energy systems operations will excel in safety and reliability. SR-2: Generation IV nuclear energy systems will have a very low likelihood and degree of reactor core damage. SR-3: Generation IV nuclear energy systems will eliminate the need for off site emergency response.

EC-1: Generation IV nuclear energy systems will have a clear life-cycle cost advantage over other energy sources. EC-2: Generation IV nuclear energy systems will have a level of financial risk comparable to other energy projects.

3. GENERATION IV REACTOR CONCEPTS

Generation IV reactor concepts were identified via a formal DOE "Request for Information" (RFI) issued in April 2001.

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3.1. Generation IV Water Cooled Reactor Concepts

This process resulted in submittal of 30 advanced water reactor concepts by researchers and industry experts in , , , , Italy, , Korea, and the U.S. In addition, the technical working group itself developed information on eight concepts, yielding a total of 38 concepts for evaluation. The technical working group consolidated all but one of the 38 reactor and fuel cycle concepts into nine distinct concept sets, based on their key common features.

3.1.1. Integrated Primary System Reactors

These light water reactor concepts are characterized by a primary system that is fully integrated in a single vessel, which makes the nuclear island more compact and eliminates the possibility of large releases of primary . The emphasis is on utilization of existing LWR technology, modularity, elimination of accident initiators, and passive systems to cope with the consequences of accident events. Of the three major Generation-IV high-level goals, this class of reactors mainly addresses the potential for superior safety and good economics. On the other hand, resource utilization and proliferation resistance are rated as comparable (or just slightly better) than current LWRs with similar fuel cycles. At this point the key R&D issues for these systems appear to be the economic viability of a modular design approach as well as the reliability and design of the in-vessel components.

3.1.2. Advanced Loop Pressurized Water Reactors

The common innovative characteristic of these reactor designs is the use of a safeguard vessel (or series of vessels and pipes) that envelopes the whole primary system for mitigation of primary system component failure. Moreover, the adoption of the additional vessel enables elimination of somesafety systems. This reactor concept offers potential for superior safety compared with the reference LWRs. However, issues to be resolved include reliability and maintenance of the primary system components that are not easily accessible, and impact of the additional vessel on the capital cost.

3.1.3. Simplified Boiling Water Reactors

These are various size boiling water reactors with natural circulation in the core region, no re- circulation pumps, and, in most cases, highly passive decay heat removal systems. With one exception (the SMART concept) the concepts within this group are all founded on existing and proven BWR technology and do not need extensive R&D for their deployment. They feature various design improvements intended to provide economic or other advantages. At this point the key R&D issue for these systems appears to be the demonstration of their economic values relative to other designs.

3.1.4. Pressure-Tube Reactors

This concept set is based on the CANDU design. The emphasis is on improving the economics by: replacing the heavy water coolant with light water; moderately increasing the thermal efficiency; simplifying and reducing the size of the nuclear island, and use of fuel.

3.1.5. Supercritical Water Cooled Reactors

The unique thermo-physical properties of supercritical water offer potential for designing nuclear reactors with significantly higher thermal efficiencies (40-45% versus the current 33-34%) and considerable plant simplification, compared to the ALWR. However, to make such systems technologically feasible:

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 498 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ • advances are required in high-temperature materials to improve corrosion, stress corrosion cracking, and wear resistance; • in neutronics to improve fuel-cycle versatility with these advanced materials; and • in neutronics and thermal-hydraulics to insure an acceptable level of safety and stability.

3.1.6. High-Conversion Reactors

These are LWRs with a tighter fuel lattice and less moderator and, therefore, a fast spectrum with greatly increased fuel utilization. Since uranium resources will become scarce in a nuclear economy, this concept set might become very important in the future. However, there are key R&D issues to be addressed, including: • neutronic stability of the core (e.g., negative void reactivity coefficient); • development of appropriate fuel cladding and core internals structural materials; • demonstration of effective coolability of tight cores, and • development of suitable proliferation resistant fuel reprocessing techniques to take advantage of the increased production of .

3.1.7. Pebble-Fuel Reactors

The emphasis in this class of reactors is on passive safety: both shutdown and decay heat removal. The fuel temperatures are significantly reduced during operation. However, the fuel fabricability and reliability in a water environment needs to be demonstrated.

3.1.8. ALWRs with Thorium-Based Fuels

The significant advantages of the once-through thorium cycles with respect to proliferation resistance and waste form stability are very attractive to society as a whole, but provide little economic incentive to the current industry. The energy resource sufficiency advantage of the U- 233/Th-232 light water fuel cycle is currently out weighed by reliability and cost issues. However, further out in the future our low cost uranium supplies will become depleted and the thorium fuel cycles will eventually become cost effective.

3.1.9. ALWRS and CANDU Reactors with Dry Recycling of Spent LWR Fuel

These technologies have significant potential for reducing spent fuel volumes and increasing fuel utilization. Key R&D issues that are identified for this fuel cycle include: • cost effective fabrication processes and equipment, and • adequate solutions for capture and immobilization of the volatile fission products that are released during the recycling process.

3.2. Generation IV Gas-Cooled Reactor

A DOE RFI and team solicitations resulted in 21 reactor system concepts submitted from France, Germany, Japan, Netherlands, and the U.S. The 21 concepts were consolidated into four concept sets. The four concept sets have been qualitatively screened to assess their potential to achieve the generation IV goals. The screening used criteria developed by the Evaluation Methodology Group in support of each goal, and measured against existing advanced light water reactor designs

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 499 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 3.2.1. Gas Cooled Thermal Reactor

General features are: • reference concepts used once through LEU fuel cycles TRISO fuel -- SiC and pyrolytic fission product barriers; • graphite moderated, coolant ″Naturally safe″ designs with conductive and radiative decay heat removal; • high temperature direct Brayton cycle power conversion Increased fuel utilization and decreased HLW due to high thermal efficiency; and • significant fuel cycle flexibility within a reactor design-LEU once-through, Pu-MA single recycle, W-Pu, HEU, Th-U233 converter.

3.2.2. Pebble Bed Reactor Systems (PBR)

In the Pebble Bed modular reactor the helium flows through the pebble bed and removes the heat generated by the nuclear reaction from the core. Five concepts were submitted. Reference concept with 115 MWe, 250 MWth, direct Brayton cycle and low excess reactivity (continuous on-line refueling). These concepts show promise for:

• modest gains in sustainability; • significant advance towards safety goals; and • comparable economic.

The R&D needs are: • fuel qualification at higher , fluences, and temperatures; • beyond design basis event behaviors (air and water ingress); and • Fuel manufacturing quality improvements

3.2.3. Prismatic Fuel Modular Reactor (PMR)

Five concepts were submitted. Reference concept with 286 MWe, 600 MWth, direct Brayton cycle, 850 C core exit temperature, LEU once-through fuel cycle. Fuel cycles submitted included waste transmutation, W-Pu burner, Th-U233 converter. These concepts show promise for: • modest gains in sustainability, • significant advance towards safety goals; and • comparable economics

The R&D needs are similar to PBR: • fuel performance qualification; • fuel manufacturing quality; • higher temperature vessel materials qualification; and • turbomachinery bearings.

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 500 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 3.2.4. Very High Temperature Reactor System (VHTR)

The general features of VHTR are: • >900 C coolant core exit temperature; and • prismatic core, 600 MWth, LEU once-through cycle Four concepts submitted show promise for: • gains in sustainability and flexibility; • significant advance towards safety goals; and • comparable economics.

The motivation for VHTR are: • nuclear power can offset other primary fuels in applications other than electricity; and • VHTRs may significantly reduce liquid and gaseous demands.

3.2.5. Gas Cooled Fast Reactor Systems (GFR)

Four concepts submitted, however there is no complete reference concept. Novel features of the four concepts may illuminate future development pathways. Three concepts use helium coolant, one uses CO2. May allow recycle and passive decay heat removal. They show promise for: • significant advance in sustainability; • comparable safety performance; and • unclear economics

The GFR R&D needs are: • fuel, structural and core materials; • passive safety system capabilities; and • recycle techniques

3.3. Liquid Metal Reactor System Concepts

A total of 33 concepts from 8 countries submitted; 27 of which were grouped into five sets with power from 75 MWe to 1500 Mwe. Fuel cycle technology in the great majority of cases was the ″pyroprocess″ (i.e. electrometallurgical technology) or the ″advanced aqueous process″ both will require extensive development.

Fuels: • mixed oxide (reference or backup in 10 concepts); • metal (16 concepts); and • nitride (6 concepts) : • (11 concepts); and

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Five groups are: • A: medium-to-large sodium-cooled, mixed-oxide fueled reactors with advanced aqueous reprocessing and pellet or vibratory compaction fabrication (5 concepts); • B: medium-to-large sodium-cooled, metal-fueled (U-Zr metal) reactors with electrochemical fuel cycle technology (pyroprocessing) (6 concepts); • C: medium-sized Pb or Pb-Bi cooled; MOX or Th-U-Zr metal alloy fueled reactors (one concept had nitride fuel); pyroprocess fuel cycle for the metal-fueled concepts, advanced aqueous or nspecified ″dry″ process for the ceramic fueled concepts. (9 concepts); • D: small, Pb or Pb-Bi cooled; metal or nitride fueled reactors with long-life ″cartridge″ or cassette cores. Fuel cycles vary. (4 concepts) • E: sodium-cooled concepts that eliminate the traditional secondary sodium loops by development of novel new steam generators. (3 concepts).

Liquid Metal Reactor Systems show promise for:

• safety, general attempt to rely on inherent safety features; • uranium resource utilization in a category by itself compared to all thermal systems; • significant waste volume reduction relative to ALWR once-through, but the key benefit would derive from meeting the widely-adopted goal of 99.9% recycle of all ; • many of the systems claim immunity (i.e. no fuel damage); • proliferation, resistance evaluation is challenging; and • economics, the great challenge, being approached through simplification of both reactors and fuel cycle facilities.

3.4. Non-Classical Nuclear System Concepts

Non-Classical reactor concepts feature higher potential to meet or exceed Generation IV performance goals. A total of 32 concepts gathered, among them 28 meet the Generation IV requirement of fission based self sustained criticality. These concepts show promise for: • significant advances can be made in conversion efficiency, diversification of energy products, resource utilization and waste minimization; • excellent non-proliferation characteristics due to one to two orders of magnitude lower fuel inventory and buildup; • minimized source term due to online separation and removal of fission products and ultralow equilibrium concentration of minor actinides; • gas/vapor core reactors could potentially eliminate the need for Offsite Emergency Planning, which is a key safety goal for the Generation IV reactors; and • many technology challenges; high temperature materials, energy conversion, dynamics and control, remote operation, fuel chemistry and fuel handling, fission product separation, and safety.

Based on the primary design features , 6 ″concept sets″ are elected:

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Molten salt cores are: • HERACLITUS, circulating fuel, natural thorium molten salt; • MSBR, Molten Salt Breeder, liquid uranium and thorium fluorides; and • AMSTER, Actinides Molten Salt Transmuter. Liquid Metal Core are: • LM-FR, Liquid Metal Equilibrium Fast Reactor, Mg-Pu Eutectic; and • MSBR, Molten Salt Breeder, liquid uranium and thorium fluorides.

3.4.2. Gas Core Concepts

Gas cores are:

• GCR/VCR-MHD, UF4 with either KF vapor Rankine cycle or He Brayton cycle. Efficient MHD energy conversion with fission enhanced ionization; • GCR/Graphite wall, neutralizes high temperature wall corrosion, and

• Plasma/Vortex flow, varieties of vortex flow GCR’s, high temperature, diverse uses, UF6 or U vapor with He or Argon.

3.4.3. Non-Conventional Cooled Reactor Concepts

Non-conventional cooled reactor concepts are: • AHTR, Advanced High Temperature Reactor with graphite matrix, molten salt cooled, high temperature diverse uses; • OCR, Organic Coolant Reactors with cheaper efficient cooling, reduced costs, and • FSEGT, Sodium Evaporation Fast Reactors, sodium evaporation cooling, and unique sodium vapor gas turbines.

The main characteristics of Molten Salt Cooled Reactors are: • significant advances can be made in conversion efficiency, and diversification of energy products; • high temperature operation at low pressure, low power density, high heat capacity; and • high temperature materials, fuel design, molten salt to water heat exchanger, mixed nuclear/hydrogen safety issues.

The main characteristics of Organic Cooled Reactors are: • high conversion ratio, superior coolant properties, low pressure operation,lower cost coolant (compared to CANDU), and • fuel (UC) reaction with water and air, coolant flammability, coolant fouling, coolant radiolysis, reactivity coefficients.

Despite many technology gaps and data uncertainties, there is no lack of innovation and revolutionary ideas in Non-Classical reactor concepts.

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4. THE FINAL GENERATION IV SYSTEMS SELECTED

The 10 Generation IV International Forum (GIF) countries have selected six concepts (from 114 submited concepts) to develop in order to meet the technology goals for new nuclear systems. GIF countries believe developing these concepts will achieve long-term benefits so nuclear energy can play an essential role worldwide. Generation IV nuclear energy systems are future, next-generation technologies that will compete in all markets with the most cost-effective technologies expected to be available over the next three decades. Comparative advantages include reduced capital cost, enhanced nuclear safety, minimal generation of nuclear waste, and further reduction of the risk of weapons materials proliferation. The purpose of Generation IV is to develop nuclear energy systems that would be available for worldwide deployment by 2030 or earlier.

The selected concepts are: • Gas-Cooled Fast Reactor (GFR), features a fast--spectrum, helium-cooled reactor and closed fuel cycle; • Lead-Cooled Fast Reactor (LFR),features a fast-spectrum lead of lead/bismuth eutectic liquid metal-cooled reactor and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides; • (MSR), produces fission power in a circulating molten salt fuel mixture with an epithermal-spectrum reactor and a full recycle fuel cycle; • Sodium-Cooled Fast Reactor (SFR), features a fast-spectrum, sodium-cooled reactor and closed fuel cycle for efficient management of actinides and conversion of fertile uranium; • Supercritical-Water-Cooled Reactor (SCWR), a high-temperature, high-pressure water-cooled reactor that operates above the thermodynamic critical point of water • Very-High-Temperature Reactor (VHTR), a graphite-moderated, helium-cooled reactor with a once-through uranium fuel cycle

4.1. Gas-cooled fast reactor (GFR)

The basic characteristics of GFR concept are:

• He (or CO2) coolant, direct Brayton cycle gas-turbine; • 850°C outlet temperature;

• 600 MWth/288 MWe; • U-TRU ceramic fuel in coated particle, dispersion, or homogeneous form; ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 504 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ • block, pebble, plate or pin core geometry; • waste minimization, and • efficient electricity generation

Fig. 1. Gas-cooled fast reactor (GFR)

The reactor physics issues: • core configuration dependent; • neutron streaming; and • data for actinides and fuel matrix candidate materials.

4.2. Lead-Cooled Fast Reactor(LFR)

The main characteristics are • Pb or Pb/Bi coolant; • LFR is cooled by natural convection with a reactor outlet coolant temperature of 550 degrees C, possibly ranging up to 800 degrees C with advanced materials;

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 505 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ • U-TRU nitride or Zr-alloy fuel pins on triangular pitch; • 120–400 Mwe; • 15–30 year core life; • core refueled as a cartridge; • transportable core; and • passive safety and operational autonomy.

Fig. 2. Lead-Cooled Fast Reactor(LFR)

Reactor physics issues: • data for actinides, Pb, Bi; • spectrum transition at core edge; and • reactivity feedback coefficients.

4.3. Molten Salt Reactor (MSR)

The main characteristics are:

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 506 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ • molten fluoride salt fuel, epithermal spectrum; • 700–800°C outlet temperature; • 1000 Mwe; • low pressure (<0.5 MPa); • circulating actinide-bearing fuel; • graphite core structure to channel flow; • actinide consumption; and • avoids fuel development and fabrication.

Fig. 3. Molten Salt Reactor (MSR)

Reactor physics issues • evolution of mobile-fuel composition; • modeling of nuclear, thermal, and physio-chemical processes; • delayed neutron precursor loss; and • resonance interference effects.

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 507 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 4.4. Sodium-Cooled Fast Reactor (SFR)

The main characteristics of SFR are:

• sodium coolant, Tout≈550°C; • 150 to 1500 Mwe; • U-TRU oxide or metal-alloy fuel; • hexagonal assemblies of fuel pins on triangular pitch; • homogenous or heterogeneous core; • consumption of LWR discharge actinides; and • efficient fissile material generation.

Fig. 4. Sodium-Cooled Fast Reactor (SFR)

Reactor physics issues • actinide data; • full-core transport effects;

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 508 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ • spectral transition at core periphery and beyond; and • accurate modeling of expansion feedback.

4.5. Supercritical-Water-Cooled Reactor (SCWR)

The supercritical water is interesting and not yet deeply studied phenomena in nature. Critical parameters for the water are 374°C, 22.1 MPa. The SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once- through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers is eliminated. The core rated thermal power is 3,575 MW resulting in a pressure vessel of 5.3 m inner diameter and 46 cm thickness in the beltline region. The vessel operates at 280 C and traditional LWR low- alloy steels such as SA-508 can be used. Because of its large size the vessel cannot be manufactured in the U.S., but appears to be within existing manufacturing capabilities in Japan. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired , a large number of which is also in use around the world. The issue of transport of coolant activation products to the balance of plant was also evaluated at INEEL. It was found that the 16N activity in the SCWR steam is about twice that in the steam of a BWR with hydrogen water chemistry. However, a simple gamma attenuation model showed that this results in shielding requirements for the SCWR only up to 12% higher than for the BWR. Moreover, because of the higher SCWR electric power, the specific shielding costs ($/kWe) associated with 16N are expected to be similar to or better than the BWR’s

The basic characteristics are: • water coolant at supercritical conditions (~25 MPa); • 510°C outlet temperature; • 1700 Mwe;

• UO2 fuel, clad with SS or Ni-based alloy; • square (or hex) assemblies with moderator rods; • high efficiency, compact plant; and • thermal or fast neutron spectrum.

Reactor physics issues: • similar to BWR’s; • increased heterogeneity; • strong coupling of neutronics and thermo-hydraulic; and • neutron streaming.

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 509 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████

Fig. 5. Supercritical-Water-Cooled Reactor (SCWR)

4.6. Very-High-Temperature Reactor

The main characteristics are: • He coolant, direct cycle; • 1000°C outlet temperature;

• 600 MWth; • coated particle fuel; • solid graphite block core; • high thermal efficiency; • hydrogen production; and • passive safety.

Reactor physics issues: • fuel double heterogeneity;

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 510 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ • stochastic behavior of pebble movement (for PBR variant); and • graphite scattering treatment.

Finally, highly efficient carbon-free means of producing hydrogen by coupling advanced high temperature nuclear reactors with thermo-chemical processes is on the horizon.

Fig. 6. Very-High-Temperature Reactor

5. NEW DESIGN AT WESTINGHOUSE: INTERNATIONAL REACTOR INNOVATIVE AND SECURE (IRIS)

IRIS is not a Generation IV design since it will be available for deployment decades ahead of the 2020 to 2030 time frame projected for the six selected Generation IV systems [5]. IRIS, however, has been the first to formulate and implement the philosophy that next generation systems should leverage their design and operational characteristics to prevent accidents to the highest extent possible. Although still at the preliminary design stage, the IRIS nuclear power plant design has moved rapidly from idea to viable commercial entry. IRIS is a modular pressurized water reactor with an integral configuration (all primary system components pumps, steam generators, pressurizer, and control rod drive mechanisms are inside the reactor vessel).

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 511 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ It is offered in configurations of single or multiple modules, each having a power rating of 1000 MWt (about 335 MWe). IRIS is a very innovative reactor design with many attractive new features, especially in the safety area, but at the same time its technology is grounded on well-proven and universally familiar water reactors experience. The main characteristics of the IRIS is integral configuration. All the main primary system components (core with reflector/shield, pressurizer, reactor coolant pumps, steam generators, and control rod drive mechanisms) are located inside the reactor pressure vessel.

Reactor vessel

The integral arrangement eliminates all the pressure vessels outside the reactor vessel, as well as the large connecting loop piping between them, resulting in a compact, more economical configuration and in the physical elimination of the LOCAs. Because the IRIS integral vessel contains all the RCS components, it is larger than a traditional RV, and has an inside diameter of 6.2 m and an overall height of 21.3 m, including the closure head (figure 7).

Steam generators

The IRIS SGs are of a once through, helical-coil tube bundle design, with the primary fluid outside the tubes. Eight steam generator modules are located in the annular space between the core barrel (outside diameter 2.85 m) and the reactor vessel (inside diameter 6.2 m). The enveloping outer diameter of the tube bundle is 1.64 m. Each SG has 656 tubes, and the tubes and headers are designed for the full external RCS pressure. A unique aspect of the IRIS SG design is that the high pressure primary coolant flows on the outside of the tubes. Thus, the IRIS SG tubes are in compression, and therefore, tensile stress corrosion cracking which, according to EPRI data, has been responsible for more than 70 percent of all the SG tube failures is automatically eliminated.

Reactor coolant pumps

The IRIS pumps are of the spool type, which has been used in marine applications and designed for chemical plant applications requiring high flow rates.

Control rod drive mechanisms

The integral configuration is ideal for locating the CRDMs inside the vessel, in the region above the core and surrounded by the steam generators. Their advantages are in safety and operation

Pressurizer

The IRIS pressurizer is integrated into the upper head of the reactor vessel. By utilizing the upper head region of the reactor vessel, the IRIS pressurizer provides a very large water and steam volume, as compared to plants with a traditional, separate, pressurizer vessel. The IRIS pressurizer has a total volume of about 71 m3, which includes a steam volume of about 49 m3 . This steam volume is about 1.6 times bigger than the pressurizer steam space for a large PWR, while IRIS has about 1/3 the core power. Because of this large steam volume-to-power ratio (about five times the value of a typical PWR), IRIS does not need a pressurizer spray function to prevent the pressurizer safety valves from lifting for any design basis heatup transients.

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 512 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Adoption of an integral configuration has a very positive impact on the reactor's overall intrinsic safety, well beyond the obvious elimination of the large LOCAs. This has allowed IRIS to implement an extremely effective "safety-by-design" approach (figure 8).

Safety-by-design approach

Of the eight class IV accidents that must be considered in PWRs, only one remains unaffected in IRIS (Design basis fuel handling accidents). All the others are either eliminated outright or are downgraded to a lower classification. This has very important implications on the IRIS approach to licensing as it will be seen later.

Fig. 7. IRIS integral layout

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 513 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ An example of the innovative thinking behind the IRIS safety-by-design approach is given by the handling of small-break LOCAs, which historically have been most plaguing to PWRs. The IRIS approach is to limit and eventually stop the loss of coolant from the vessel rather than to rely on active or passive systems to inject water into the RV. Advantage of the following three features of the design: • the large coolant inventory in the reactor vessel; • an emergency heat removal system (EHRS) employing the steam generators to remove heat directly from inside the RV, thus depressurizing the RV by condensing steam, rather than by discharging mass;

Fig. 8. IRIS spherical steel containment arrangement

• the compact, small-diameter, high design pressure containment , which during the accident becomes thermodynamically coupled with the vessel and assists in limiting the blowdown from the RV by rapidly equalizing the vessel and containment pressures. The IRIS small spherical containment has a design pressure more than three times the value typical of loop PWR containments, at the same shell thickness; and • innovative containment design, which practically eliminates small-to-medium loss-of-coolant accidents (LOCAs) as a safety concern. The internal configuration, however, yields a containment much smaller than those for conventional PWRs. IRIS employs a spherical containment, with a

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 514 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ diameter of 25 m, slightly more than half the diameter of the cylindrical containment for a 600- MWe PWR.

6. NEW DESIGN AT UC BERKELEY: ENCAPSULATED NUCLEAR HEAT SOURCE (ENHS)

The Department of Nuclear Engineering at University of California at Berkeley, in collaboration with Westinghouse and LLNL designed the Encapsulated Nuclear Heat source (ENHS) [6]. This design is based on Russian lead-bismuth coolant technology.

The main characteristics of the ENHS are: • new fission core designs that can burn long-lived waste at the same rate it is created;

• 125 MWth, Pb-Bi cooled, fast spectrum; • factory-fabricated, transportable; • 20 year refueling module; • inserted in secondary Pb-Bi pool at clients site; • natural circulation heat transport across module wall; • semi-autonomous operation; • potential marketable in developing countries with developing grids; • shipped from factory in frozen Pb-Bi; • used module returned to factory; • passive safety; and • passive decay heat removal.

7. RESULTS AT UC BERKELEY AND VINČA INSTITUTE

IRIS benchmark calculations

A new methodology SAS2H/KENO-V.a [7] for fuel depletion analysis of entire cores of the IRIS reactor, founded on the application of SCALE-4.4a [8] sequences was developed. This methodology combines a 3D Monte Carlo full core calculation of node power distribution and a 1D Wigner-Seitz equivalent cell transport method for independent depletion calculation of each of the nodes. In order to assess the feasibility/practicality of the SAS2H/KENO-Va methodology for full core modeling, we compared it with the well-established Westinghouse ALPHA/PHOENIX/ANC deterministic code system [9]. In analyzing the IRIS core benchmark, it was not always possible/practical to use exactly identical assumptions. Therefore, some differences in the results are expected. We should emphasize that the objective of this comparison was not to investigate and resolve these small differences, but to establish that with reasonably similar assumptions we can obtain reasonably close results. The reasons for the differences include the following: (1) non-standard use of the Westinghouse codes, (2) slightly different reflector representation, (3) different treatment of the Doppler feedback (the fuel temperature feedback has been completely removed in SAS2H/KENO- Va, but not completely eliminated in ANC), (4) somewhat different treatment of fission products close

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 515 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ to BOC, (5) and change in fuel isotopics is followed on a different axial mesh, (6) statistical noise in MC results, combined with a possibility that the solution is still not completely converged to fundamental mode, (7) different cross section libraries, including B-V and B-VI based data, (8) possibly some differences in material compositions, densities (different modeling of expansion), grid volume, and temperatures. It should also be noted that the Benchmark#44 core configuration does not necessarily represent an actual (acceptable) core configuration and that some non-physical conditions are imposed (e.g., some feedback effects are intentionally removed).

Fig. 9. ENHS integral layout

Figure 10 compares keff . The solid line represents PHOENIX/ANC, the dashed line is SAS2H/KENO-V.a. The difference is acceptable, almost constant over the whole range, and primarily due to different assumptions. Figure 11 compares power peaking factors as a function of burnup; the behavior is very similar. SAS2H/KENO-Va results are always slightly higher, typically about 6 percent, with additional several percent variability. This average difference is consistent with the difference in applied Doppler feedback, while the variability is partly due to MC statistics. The 2σ uncertainty in SAS2H/KENO-V.a results at the maximum power density location is typically ~2%,

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 516 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ and higher elsewhere, contributing some of the variability in the observed differences. Overall, a very good agreement in power peaking factors is observed.

Figure 12 compares axial power distributions at 150, 6650, 18650, and 36650 MWd/tHM, where the PHOENIX/ANC results have been projected to the SAS2H/KENO-V.a axial mesh. The observed agreement is very good for the main part of the cycle. The difference is somewhat higher at EOC; this is tentatively explained by the differences in modeling, i.e., different axial mesh used by the two code systems. Consequently, fuel isotopics is followed for axial zones of different length, this difference builds up with fuel depletion, and for the very long cycle considered here eventually impacts the axial power shape. In summary, a very good agreement of the two code system results has been observed. Moreover, the existing differences may be attributed in large part to different modeling assumptions.

1.40 5.00

SAS2H/KENO-Va SAS2H/KENO-Va 1.30 PHOENIX/ANC PHOENIX/ANC

4.00 1.20

1.10 3.00 K-eff Fq

1.00

2.00 0.90

0.80 1.00 0 5000 10000 15000 20000 25000 30000 35000 40000 0 5000 10000 15000 20000 25000 30000 35000 40000 BURNUP (MWd/tU) BURNUP (MWd/tU)

Figure 10. keff evolution with burnup Figure 11. Evolution, with burnup, of core peak to average power density ratio

3 3

2 2

1 1 RELATIVE POWER RELATIVE POWER RELATIVE

0 0 0 60.96 121.92 182.88 243.84 304.8 365.76 426.72 0 60.96 121.92 182.88 243.84 304.8 365.76 426.72 AXIAL POSITION (Z) AXIAL POSITION (Z)

(a) 150 MWd/tU (b) 6650 MWd/tU

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 517 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████

3 3

2 2

1 1 RELATIVE POWER RELATIVE POWER RELATIVE

0 0 0 60.96 121.92 182.88 243.84 304.8 365.76 426.72 0 60.96 121.92 182.88 243.84 304.8 365.76 426.72 AXIAL POSITION (Z) AXIAL POSITION (Z)

(c) 18650 MWd/tU (d) 36650 MWd/tU

Figure 12 Axial power profile at representative core burnups

ENHS benchmark calculations

Two computational procedures for the analysis of the ENHS benchmark core. The first procedure is based on the application of the MCNP -4C [10] and ORIGEN2.1 [11] codes, interfaced by the MOCUP [12] driver were investigated. The second procedure KWO2 [13], recently developed by the authors of this paper, is based on the coupling of the KENO -V.a and ORIGEN2.1 codes. These utility codes are designed for the reference calculations of long-life cores in 2D or 3D geometry models The results obtained with these procedures using various cross section libraries are compared with the results obtained by using the fast reactor neutronics design tools [14] in use at Argonne National Laboratory (ANL).

The used continuous energy libraries, i.e., the VMCCS (from 1998) and ENDF60 did not provide for the self-shielding effects in the resolved resonance energy region, although MCNP-4C was designed to include this effect by using the probability table method (the most rigorous method for the treatment of the unresolved resonance self-shielding effect in the Monte Carlo calculations). In order to examine the self-shielding effect in the unresolved resonance energy range, the single zone model of the ENHS benchmark core was analysed with the new VMCSS libraries based on the ENDF/B-VI.7 and ENDF/B-V.2, equipped with probability table (ptable) data. On the other hand, it was recognised that even the 238-group SCALE-4.4a library is not good enough for accounting of this effect. That is why we tried to use the best option for modelling of resonance self-shielding effect in the ENHS benchmark core with CSAS25 module. We found that the best option for accounting of these effects in KENO-V.a calculation is to use the new CSAS25 module from SCALE-5.0 code system to prepare both: the 1D matrix of the multigroup self-shielded cross section and the 2D matrix of elastic scattering cross sections by using the continuous energy neutron flux calculated in the simplified 1D geometrical models of ENHS benchmark core.

The final results for the single zone model of ENHS benchmark core, obtained with both improved MOCUP and KWO2 procedures are given in Figure 4. As can be seen from Figure 4, a good agreement was obtained between ENDF/B-VI and ENDF/B-V based MCNP-4C results, and also between MCNP-4C and SCALE-5.0 results based on the ENDF/B-V.2 evaluation during more that 20 years of core cycle.

████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ 518 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████ Fifth Yugoslav Nuclear Society Conference YUNSC - 2004 ████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████████

1.03

1.02

1.01

1.00

0.99

MOCUP (2D-rz, Continuous Energy /B-VI.8, 30A+95FP) 0.98 MOCUP (2D-rz, Continuous Energy /B-V.2, 30A+95FP) KWO2/ver-5.0) (2D-rz, 238 groups /B-V.2, 25A+95FP) Effective neutron multiplication factor multiplication Effective neutron REBUS-3 (2D-rz, 230 groups /B5-V.2, 25A+5PFP) 0.97 0 5 10 15 20 25 30 Irradiation time (Years)

Figure 13. Comparison of keff calculations of single zone model of the ENHS benchmark core

8. CONLUSION

An overview of the status of the Generation IV design has been provided, with particular emphasis on the integral layout of the reactor coolant system and on the innovative approach to safety. The integral layout offers very significant advantages in terms of performance, simplicity, and compactness. It has been demonstrated that it has an extremely positive impact on the overall reactor safety response to postulated accidents. The net result is a design with significantly reduced complexity, improved operability, and extensive plant simplifications.

ACKNOWLEDGMENTS

This work was supported by the US Department of Energy NERI program under contracts No. DE-FG03-99SF21955 and FG03-99SF21889 and by the Ministry of Science and Environment Protection of Serbia under Contract No. 1958 ("Transport processes of particles in fission and fusion systems").

REFERENCES

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