Publications

A High Efficiency and Environmentally Friendly Nuclear Reactor (HEER) for Electricity and Hydrogen

M.S. Kazimi, P. Hejzlar, Y. Shatilla, Bo Feng, Yu-Chih Ko, E. Pilat, K. Shirvan, J. Whitman, and A. Hamed

MIT-ANP-TR-125 (October 2009)

Executive Summary

This project has examined approaches for the design of nuclear power plants to address future needs for electricity, potable water and high temperature heat through a highly efficient and environmentally friendly reactor (HEER). HEER reactors provide high power output and energy conversion efficiency, with minimum production of spent fuel and waste. These advanced reactors utilize innovative fuel designs to enable efficient fuel utilization; they heat the coolant to a higher temperature than typical water cooled reactors using an integral vessel for the reactor and steam generation in order to maximize the power plant efficiency and safety. The spent fuel production rate is reduced by a factor of about 2 in comparison to today’s reactors in Europe and Japan. Finally, to be ready for deployment in the Middle East, the new reactor concept exhibits an added level of proliferation resistance, through less frequent refueling and a higher level of fuel burnup, hence radioactivity, at discharge.

The project examined the use of several novel reactor designs to achieve its goals using watercooled or salt-cooled reactors, over two years. During the first year of the project, three concepts were identified for investigation: A) Liquid Salt Thermal Reactor for high thermal efficiency,; B) Superheat Boiling Water Reactor for high thermal efficiency; and C) a modified International Reactor Innovative and Secure (IRIS) design for smaller reactor vessel. During the second year, conceptual designs of all three concepts were developed to demonstrate their feasibility and provide their primary operating characteristics.

The Liquid Salt Thermal Reactor (LSR) is a new reactor concept that uses uranium zirconium hydride, U0.31ZrH1.6, fuel and leverages recent developments in silicon carbide cladding to allow core cooling by fluoride liquid salt (NaF-BeF2). In contrast to traditional thermal reactors, which operate with coolants at high pressure but modest temperatures, the LSR operates at atmospheric pressure, reducing the potential and consequences of a primary system breach, but at much higher temperatures that allow a large increase in thermal efficiency to nearly 45%. A 4m high reactor core yielded an acceptable core barrel diameter of 4.5m for the 1000 MWth reactor. The reactor core itself is about the same size as today’s PWR cores. Thermal-hydraulic optimization for the primary side of the reactor, as well as simplified analysis of the secondary side led to a total reactor output of 448 MWe, with an 8.75m diameter, 15.75m long vessel. The design limits fast neutron fluence at the vessel to acceptable levels. The design uses an in-vessel intermediate heat exchanger (IHX) with 1.5cm outer diameter tubes with a pitch to diameter ratio of 1.15. Pumps were also sized. Neutronic aspects of the individual fuel assembly and the total core designs were determined using MIT’s MCODE system that combines the MCNP monte-carlo code and the ORIGEN depletion code. It was shown that uranium zirconium hydride fuel rods enriched to slightly below the non-proliferation limit of 20% in a hexagonal lattice can achieve a core life of more than 10 years without refueling, this being a significant addition to proliferation-resistance. The well-known gadolinia burnable poison is utilized to help control reactivity over this long time span.

Nuclear superheat is another attractive approach to appreciably increase the thermal efficiency of boiling water reactors (BWRs), reduce the cost of generating electricity, and improve fuel utilization while reducing spent fuel per unit of generated energy. However, a superheat BWR faces a number of design challenges (in fuel element design, reactor stability and materials reliability) that need to be overcome before a successful concept can be developed and deployed. After reviewing previous designs and concepts, as well as the operating experience of the few nuclear power plants that had been built and operated with integral superheaters, a design for an Annular-fueled Superheat Boiling Water Reactor (A-SBWR) was developed. It is an innovative design applying the nuclear fission energy to achieve generate superheated steam at a temperature of 510°C or higher, resulting in a power conversion efficiency of more than 42%. This compares favorably with existing light water reactors, which provide only saturated steam and therefore have efficiencies on the order of only 33%. The A-SBWR is applicable to any reactor power level, although the example in this report is given for a medium power unit of 500 MWe. It employs annular fuel elements cooled internally by steam and externally by boiling water. This feature avoids the insulation and uneven moderation problems, which were among the challenging parts for the older superheat reactor concepts. Thermal design for the A-SBWR, included selection of an appropriate power density based on evaluations of the critical heat flux, the maximum cladding and fuel temperatures, and the maximum achievable steam superheat. A power density of 48.2 (kW/L) was adopted, only slightly below that of the reference BWR; and it leads to a maximum fuel meat temperature of 1099°C in the hot channel, much lower than the corresponding temperature of a typical solid fuel rod in the reference BWR. It provides a steam outlet temperature above 520°C for the average channel, yielding significant superheat; without violating the critical heat flux constraint or the cladding temperature constraint. Steam velocities in the hot channel range from 35 to 67 (m/s). An 11 step conceptual startup procedure was then laid out with diagrams showing the locations of water and steam in the various parts of the primary system during the course of the startup. A preliminary neutronic analysis for the proposed design was initiated with a single pin benchmark study normalizing the CASMO deterministic code to be used for core neutronics studies by comparison with the MCNP monte carlo code. The deterministic code was then used to evaluate the initial nuclear reactivity of the proposed A-SBWR fuel assembly. This was found to be somewhat less than that of the reference BWR assembly, because the T91 cladding material used in the A-SBWR has a higher neutron absorption cross section than the Zircaloy used in the reference BWR, and also because the annular fuel of the A-SBWR, with its inner as well as outer surfaces, requires more cladding material.

The International Reactor Innovative and Secure (IRIS) reactor, an existing concept developed by an international coalition of industrial and academic entities with coordination by Westinghouse, is based on a typical pressurized water reactor, but differs in two important ways: first, the core design results in a refueling interval of five to ten years compared to the one to two year interval characterizing typical light water reactors; second, the component layout is modified from a loop design to a “pot” design that includes steam generators, coolant pumps and pressurizer inside the reactor vessel. While the latter change enhances the safety characteristics, it unfortunately requires a much larger reactor vessel than would be needed for a typical PWR of the same power rating. To increase the power density and decrease the vessel size, the HEER project designed a new more compact IRIS version to reduce the physical size of the unit and the amount and cost of construction material required, and facilitate shipping of the vessel. The HEER developed conceptual compact IRIS design utilizes printed circuit heat exchangers (PCHE), and intends to use higher power density core as well. Extensive work on PWR’s high power fuel using annular fuel has been done at MIT, but two remaining issues had to be resolved: 1) mitigation of the reduced shutdown margin due to higher uranium enrichment and 2) evaluation of the impact of the fuel’s inner channel flow constriction due to possible blockage and oxide growth. Both of these issues were resolved during the first year. Specifically, the reactivity shutdown margin of PWRs using this fuel was increased to a sufficient amount by changing the control rod material from Ag-In-Cd to B4C containing enriched boron. The thermal-hydraulic performance of the fuel while subject to 35-40% inner channel constrictions due to partial entrance blockage and crud/oxide growth was evaluated and found to be acceptable. The use of compact printed circuit heat exchangers (PCHE) in place of the traditional coil steam generators (OTHSG) was found to allow a 26% reduction in vessel diameter from 6.1 to 4.5m. Use of the PCHE instead of OTHSG yielded additional advantages: a volume reduction of 48 per cent per steam generator; a secondary side pressure drop smaller by 200 kPa, which results in 3 degrees more superheat than the OTHSG; a heat exchanger height reduction of 4 meters, resulting in the PCHE being at higher elevation than the OTHSG and thus increasing natural circulation power. The main concern arising from using PCHE is the large reduction in the volume of water present in the heat exchanger. Also noteworthy is the increase in heat exchanger power density by a factor of 93. The much smaller heat exchange volume enables increasing the total SG capacity thus allowing an increase of the total power output without enlarging the vessel volume, or alternatively, it could allow more redundancy in the heat exchanger capacity.

The three conceptual designs characterized in this report show promise but are far from proven as reliable technology options. Each reactor has yet to undergo sufficient materials and component testing. Furthermore, no transient analyses have been undertaken. Thus, considerable work is needed to make these reactors a reality. In terms of the level of development needed, the compact IRIS reactor is least demanding, followed by the superheat reactor. The molten salt cooled reactor requires the most effort in development, and should be only be pursued if high temperature applications are a priority.

The technology of desalination was reviewed and its implications for the required operating conditions of nuclear power reactors were determined. In addition related aspects, like environmental impact and a comparison with alternative heat sources for desalination, have been summarized. Desalination technology has been well established since the mid-20th century and is widely deployed in the Middle East and North Africa. Indeed, it plays a vital role as a part of the strategic solution to the fresh water shortage problem that is becoming a question of life in many areas of the world. Generally, both distillation and membrane processes could be coupled with nuclear reactors, but distillation processes will require more caution. Nuclear desalination is a feasible option which has been demonstrated in Japan and Kazakhstan. There are more challenges to be overcome, but the main concern is to further reduce the cost and to secure a safe coupling scheme and operation mode. Desalination cost has considerably decreased over the last three decades and desalination technology, particularly reverse osmosis, may compete soon with conventional potable water supplies even in the non-water stressed region. Different cost reduction techniques were reviewed, including the use of hybrid systems and pre-heat reverse osmosis. Because of the seawater high salinity condition in the Arabian Gulf region, multistage flashing rather than reverse osmosis is usually needed. However, recent trends to use a hybrid approach, as was demonstrated in the Sharjah desalination plant, may be possible. Multi-effect distillation, which is a more efficient process, has the potential to replace the widely deployed multistage flashing plants in the years to come. By no means should the temperature of steam supplied to the distillation plants exceed 150°C, and in most cases it can be much less. For the reverse osmosis approach, means for pressure generation across appropriate membranes, combined with microfiltration are needed. The utilization of waste heat from power plants to preheat the feedwater, and the introduction of energy recovery devices were recommended based on recent studies.