ENCAPSULATED NUCLEAR HEAT SOURCE REACTORS FOR ENERGY SECURITY

Transcription

1 15 th Pacific Basin Nuclear Conference, Sidney, Australia, October 15-20, 2006 ENCAPSULATED NUCLEAR HEAT SOURCE REACTORS FOR ENERGY SECURITY Greenspan E 1., Hong S.G. 1,2, Monti L 1,3., Okawa T 1,4., Sumini M 3. and Susplugas A 1. 1 University of California, Berkeley, CA, USA 2 Korea Atomic Energy Research Institute, Daejon, Korea 3 Bologna University, Bologna, Italy 4 Japan Nuclear Cycle Development Institute, Fukui-ken, Japan 1. Introduction A spectrum of Encapsulated Nuclear Heat Source (ENHS) reactors have been conceptually designed over the last few years [1-10]; they span a power range from less than 100 MW th to over 200 MW th and consider a number of coolants and fuel types. Common features of all these designs include very long life cores approaching or exceeding 20 effective full power years; nearly zero burnup reactivity swing; natural circulation; superb safety; autonomous load following capability; no on-site refueling; simplicity of operation and maintenance. These features make the ENHS and the like reactors of particular interest for developing countries because of the following reasons: they can provide (a) energy to communities not connected to a central electricity grid even in countries having limited technological infrastructure; (b) energy security with minimum risk of proliferation. Following a brief description of the reference ENHS reactor design (Sec. 2) we ll elaborate on the unique features of these reactors and discuss how these features could make these reactors of particular interest for developing countries (Sec. 3). A partial summary of different ENHS design options developed so far is provided in Section 4; it considers only Pb-Bi cooled uniform fuel core designs. 2. Reference ENHS Reactor The design features of the ENHS reactor include the following: A small modular reactor; nominal power level is 125 MW th There are no fuel assemblies in the core; each fuel rod is anchored in the factory to the grid plate Nearly constant fissile fuel content and neutron multiplication factor; hence, very small excess reactivity built-in and very simple reactor control system that requires adjustment for burn-up only once every few years Nearly constant power density shape across the core throughout its life There are no pumps or valves in the primary and secondary coolant loops. The coolants flow by natural circulation The natural circulation results in passive load following capability and autonomous control. It also makes loss-of-flow accidents inconceivable There are no special decay heat removal systems other than a Reactor Vessel Air-Cooling System (RVACS) that uses natural air draft for the heat sink As all structural components exposed to neutrons are disposed of with the ENHS module once in more than 20 years, the ENHS reactor lifetime might exceed 100 years. Figures 1 and 2 show simplified schematic views of the ENHS reactor. The reference ENHS reactor has two coolant circuits, both being of a pool type; the primary coolant circulates inside the ENHS module while the secondary coolant circulates in the pool the ENHS module is inserted in. The two coolants

2 interface each other across the intermediate heat exchanger (IHX) that is an integral part of the ENHS module. This arrangement was selected for a couple of reasons: (1) There are no mechanical connections between the ENHS module and the energy conversion system; this simplifies interfacing the module with the power plant. (2) Neither the ENHS module vessel nor the IHX have to withstand large pressure differentials. This simplifies the IHX design as well as the ENHS vessel design. Schematic vertical cut through the ENHS reactor 3m 2m 30m 27m 8m 3m 2m Cross Section of Stack Number of Stacks = 4 Seismic isolators 6.94m (I.D.) 3.64m (O.D; t=0.05) Steam generators Underground silo m Reactor pool ENHS module (replaceable) Reactor Vessel Air Cooling System (RVACS) Figure 1 Schematic view of the ENHS reactor Expanded view of the ENHS reactor (not to scale) Steam generator Secondary coolant Primary coolant Heat exchanger Peripheral control assembly Central control assembly Core Figure 2 Schematic vertical view of an ENHS module and pool

3 The ENHS module is designed to be as simple, robust and proliferation resistant as possible. There are no moving components except for the control (6 pieces) and safety (1 piece) elements drives. The core is an annular cylinder made of uniform lattice of fuel rods that are individually tied-up to the lower grid plate; there are no fuel assemblies. There are no blanket and no solid reflector assemblies. The safety element is located in the coolant-filled cavity at the core center while the control elements are in the form of six segments of a 5 cm thick annulus that surrounds the core as in the 4S reactor design [11]. The ENHS modules are fabricated, fuelled and weld sealed in the factory. They are transported to the power plant site with the fuel embedded in solidified Pb-Bi. No on-site refueling or fuelling hardware is required. At end of life, the ENHS module serves as a spent fuel storage cask and, later, it can be inserted in a qualified shipping cask for return for recycling to the supplier country or to a regional fuel cycle center.. The ENHS core is fuel-self-sustaining (FSS). There is a slight build-up of fissile fuel with burn-up that is used to compensate for the negative reactivity effect of the fission products that accumulate during the core life. The fuel is discharged from the FSS core when reaching its radiation damage limit. What is necessary for reusing this discharged fuel is to remove all or part of the fission products, mix the heavy metal (HM) with makeup fuel, re-fabricate fuel elements and load them into the core of a new ENHS module. The make-up fuel, approximately 5% to 7% of the fuel loading, can be either depleted uranium or natural uranium or spent fuel from LWRs. The net result is that such a FSS can recycle its actinides many times in a highly proliferation resistant manner - without partitioning of actinides and without using uranium enrichment services. Relatively simple and proliferation resistant processes could be used for the extraction of fission products [2, 5]. 3. Energy Security for Developing Countries ENHS reactors could be of particular interest for providing electricity, thermal energy and, possibly, desalinated water and hydrogen to communities that are not connected to a central electricity grid such as to many pacific islands and to remote communities in the mainland of different countries [5, 12-13]. ENHS reactors provide energy security by virtue of the following features: (1) Once an ENHS reactor is commissioned, the community it serves in the user country has assured clean energy supply for more than 20 years without needing fuel supply. (2) The energy value of the fuel loaded (in the factory) in the ENHS module is preserved; what is needed for generating energy for additional 20+ years is to add some depleted or natural uranium for makeup fuel. The fuel recycling is envisioned done by either the supplier country or by a regional or international fuel cycle centre. (3) As the ENHS module is replaced at its entirety at the end of the core life that is brought about by radiation damage, the ENHS plant life is likely to last for over 100 years. The above features also offer exceptional stability in the price of energy generated by ENHS reactors. The ENHS reactor and its fuel cycle are designed to provide a highly secure nuclear energy system so as to reduce the complexity and expense of eliminating concerns about nuclear proliferation and severe nuclear accidents, even when the user is in the initial phase of developing an energy infrastructure. The elimination, or great reduction in proliferation concerns is a result of technical attributes of the ENHS lack of excess to the fuel and to neutrons in the user country, as well as of institutional attributes the user country can enjoy the environmental, energy security and reliability benefits of nuclear energy at a competitive cost without requiring to assume the burdens of developing its own enrichment and reprocessing capabilities. [12, 13] 4. Alternative ENHS Reactor designs The reference ENHS reactor is designed [4, 5] to have a metallic alloy fuel made of plutonium and uranium with 10 weight % zirconium of the IFR type and lead-bismuth eutectic coolant of the type used for the Russians Alpha submarine reactors. The Pu is assumed extracted from PWR spent fuel that underwent a burnup of 50 GWD/tHM and was cooled for 10 years. The core is made of a uniform composition fuel rods positioned in a triangular lattice making an annular cylinder. The cross sectional area of the central cavity is assumed equal to the combined cross-section area of 217 unit cells. At full power both the central absorber and the peripheral absorbers are assumed totally withdrawn from the core and the coolant takes their place.

4 The fuel and fission gas plenum length is 125 cm each. The clad inner radius and its thickness are 0.65 cm and 0.13 cm, respectively. The fuel smear density is assumed to be 75% of the nominal density. The choice of this combination of design parameters was dictated by a combination of clad integrity, natural circulation flow and neutronic considerations [3, 5]. The number of fuel rods in the reference core is taken to be The primary core design variables are the fuel rod pitch-to-diameter ratio (P/D) and Pu weight %; a combination of these variables is adjusted to give beginning-of-cycle k eff of slightly over 1.0 and a burnup reactivity swing that is smaller than β eff. The primary thermal-hydraulic analysis design variables are the number (and thickness) of the IHX channels; the IHX length is kept constant at 13 m. The thermal-hydraulic analysis determines the maximum power level that can be removed from the core by natural circulation without exceeding constraints on the peak clad, peak fuel and peak fuel-clad interface temperature. A number of alternative ENHS designs were worked out over the last few years [4-10], all featuring a nearly zero burnup reactivity swing and natural circulation cooling. They differ from the reference ENHS designs in the fuel type, coolant type and initial fuel composition. All three variables were found to significantly affect the power attainable from an ENHS module of a given length. They also affect the value of reactivity coefficients. Selected design and performance variables are summarized in Table 1 for full-size ENHS reactor designs and in Table 2 for reduced size designs, all designs using Pb-Bi coolants and uniform composition and dimension fuel rods. The goal of the full-size designs was to explore the effect of the fuel type and composition on the attainable power, among other characteristics. The goal of the reduced size designs was to identify the minimum power an ENHS module could deliver while preserving the reference ENHS specific power and natural circulation cooling while providing nearly zero burnup reactivity swing. All the full-size designs use the reference design number of fuel rods and fuel rod dimensions, while the reduced size designs use somewhat shorter and thicker fuel rods 115 rather than 125 cm 0.83 rather than 0.65 cm in radius. Correspondingly, the clad thickness was increased from 0.13 cm to 0.21 cm. Table 1 Selected design and performance characteristics of alternative full-size ENHS cores Characteristic Reference Nitride Pu- TRU- Pu-UZr(10) N nat Nitride N UZr(10) UZr(10) Loaded Pu source; cooling time (y) PWR; 10 PWR; 10 PWR; 10 PWR; 20 ENHS c ; 5 Pitch-to-diameter ratio Attainable power (MW th ) Pu (TRU) weight % Av. heat generation rate (W/cm) Specific power (kw/kg) Average power density (W/cc) Average discharge BU (a) (GWd/t) NA (b) NA (b) Peak discharge BU (a) (GWd/tHM) NA NA Core life at attainable power (EFPY) (d) 13.8 (d) Doppler effect (dk/kk - o C) E E E-6 NA NA Axial fuel expansion (dk/kk - o C) E E E-6 NA NA Coolant expansion (dk/kk - o C) E E E-6 NA NA Core radial expansion (dk/kk - o C) E E E-6 NA NA Voiding whole core (dk;%) + gas plenum NA NA (a) Corresponding to peak fast neutron fluence of 4E+23 n/cm 2. (b) Not Available (did not calculate) (c) From equilibrium cycle of ENHS. (d) An approximate estimate assuming spectrum as of reference case The power figures in Table 1 and 2 are based on a recently revised thermal-hydraulic analysis [14] that assumes 450 o C for primary coolant core inlet temperature, 600 o C for peak core coolant outlet temperature and 650 o C for peak clad temperature. This analysis is based on the methodology developed by Sienicki [15] that was adopted by Okawa [16]. The results of the neutronic analysis reported in the tables were obtained assuming an average linear heat generation rate of W/cm for the full size cores and W/cm for the small cores. Also, some of the neutronic analysis assumed lower core coolant average temperature (by ~50 o C) than assumed by the T-H analysis. Hence, the neutronic and thermal-hydraulic results are not fully

5 consistent; an additional iteration or two are needed to achieve full consistency. Nevertheless, the general message of this paper is not affected by the present inconsistency. Table 2 Selected design and performance characteristics of alternative reduced-size ENHS cores Performance Parameter Reference S1 SM1 SM2 SM3 MA concentration in initial fuel (wt% of Active core height (cm) P/D ratio Core outer radius (cm) Attainable power (MW th ) Average Pu or TRU concentration (wt %) Average linear heat generation rate (W/cm) Specific power (kw/kg) Average volumetric power density (W/cc) Core life at attainable power (EFPY) 16.4 (a) Peak discharged burnup (GWD/tHM) (a) Average discharged burnup (GWD/tHM) 53.1 (a) Doppler effect (dk/kk - o C) Axial fuel expansion (dk/kk - o C) E E E E E E E E-6 Coolant expansion (dk/kk - o C) E E E E-6 Grid-plate radial expansion (dk/kk - o C) E E E E E E E E-6 Voiding whole core (%dk) +gas plenum Total heavy metal inventory (kg) Total plutonium inventory (kg) Total minor actinide inventory (kg) (a) Corresponding to peak fast neutron fluence of 4E+23 n/cm Table 1 shows that the power attainable from the reference ENHS design is actually 167 MW th rather than the 125 MW th published before [1-5]; this power increase is due to improvement in the thermal-hydraulic design optimization. By increasing the cooling time of the Pu discharged from PWR from 10 to 20 years, thereby reducing the 241 Pu contents, the optimal P/D ratio significantly increases making it possible to operate the ENHS module at 200 MW th.. Even larger P/D is required for cores loaded with the fuel recycled from previous ENHS cores, thereby enabling operating the ENHS up to 230 MW th. Use of nitride fuel made of nitrogen highly enriched with 15 N also enables increasing the ENHS power output to above 200 MW th. The relatively high P/D ratio cores also feature significantly enhanced discharge burnup as a result of having softer spectrum, thus reducing the fast neutron fluence per MWD/t. Table 2 shows that by loading the ENHS core with enhanced amount of minor actinides it is possible to significantly increase the P/D ratio required for obtaining nearly zero burnup reactivity swing and, thereby, significantly increase the attainable power level. The minimum power core featuring the reference core specific power has a power level of approximately 80 MW th ; its P/D is ~1.33; that is, it is between the SM2 and SM3 designs. Additional alternative ENHS core designs were developed, including a stepped-geometry core that features negative or nearly negative void coefficient of reactivity [8] for enhanced safety, cores having variable diameter or variable enrichment fuel rods for power flattening and enhanced power generation [9] and cores featuring molten salts [10] or sodium [4, 7] coolants.

6 5. Conclusions It is possible to design ENHS modules in the power range from 80 to 230 MW th while preserving the unique features of the ENHS nearly zero burnup reactivity swing, up to the radiation damage limit of the HT-9 clad, and natural circulation. ENHS nuclear-battery type reactors can provide developing countries secure energy without having to make heavy investment in developing central electricity grids and in developing indigenous fuel cycle infrastructure. The latter feature, along with the technical proliferation resistant and superb safety attributes of the ENHS are expected to make the leading nuclear countries willing to supply ENHS-like reactors to the developing countries without concerns about proliferation and severe accidents. Acknowledgment This work was supported in part by the Lawrence Livermore National Laboratory under contract number B and in part by the Post-Doctoral Fellowship Program of Korea Science and Engineering Foundation (KOSEF). References [1] Greenspan, E., et al., The encapsulated nuclear heat source reactor - a generation IV reactor, Proc. of Global-2001, Paris, France, September [2] Greenspan, E., et al., The encapsulated nuclear heat source reactor for proliferation-resistant nuclear energy, Proc. of Global-2001, Paris, France, September [3] Greenspan, E., et al., The long-life core encapsulated nuclear heat source generation IV reactor, Proc. of the Int. Congress on Advanced Nuclear Power Plant, ICAPP, Hollywood, FL, June 9-13, 2002, paper # 1081 [4] Hong, S.G., Greenspan, E., Kim, Y.I., The encapsulated nuclear heat source (ENHS) reactor core design, Nuclear Technology, Vol. 149, No. 1, January 2005, pp [5] Greenspan, E., ENHS general information, technical features, and operating characteristics, To be published in the IAEA Status Report on Innovative Small and Medium-Sized Reactors (SMRs), 2006 [6] Hong, S.G., Greenspan, E., Kim, Y.I., Use of minor actinides for reduced power ENHS reactor core design, Proceedings of GLOBAL 05, Tsukuba, Japan, October 9-13, 2005 [7] Hong, S.G., Greenspan, E., Kim, Y.I., Alternative design options for the Encapsulated Nuclear Heat Source Reactor (ENHS), Proc. of the 2003 International Congress on Advances in Nuclear Power Plants, Cordoba, Spain, May 4-7, 2003, Paper # 3248 [8] Okawa T. and Greenspan, E., Alternative ENHS Core Design Using Stepped Geometry Core, Transactions American Nuclear Society, Vol. 91, November [9] Hong S.G. and Greenspan E., Power Flattening Options for the ENHS (Encapsulated Nuclear Heat Source) Core, Proc. Int. Symp. on Innovative Nuclear Energy Systems, Tokyo, Japan, October 2004 [10] Hong S.G. and Greenspan E., Molten Salt Cooled Encapsulated Nuclear Heat Source (ENHS)-Like Reactors, Proc. Int. Symp. on Innovative Nuclear Energy Systems, Tokyo, Japan, October 2004 [11] Hattori, S. and Minato, A., Current status of 4S plant design, Proceedings of the 2 nd ASME-JSME International Conference on Nuclear Engineering: ICONE-2, CA, March 21-24, 1993 [12] Brown, N.W., et al., Liquid metal cooled reactors and fuel cycles for international security, Proc. International Conference on Nuclear Engineering: ICONE-11, Tokyo, Japan, April 20-23, 2003, paper number ICONE [13] Wade, D.C., et al., ENHS: The encapsulated nuclear heat source A nuclear energy concept for emerging worldwide energy markets, Proc. Int. Conf. on Nuclear Engineering, ICONE-10, Arlington, VA, April 14-18, 2002, paper # ICON [14] Susplugas, E. and Greenspan, E., Implementation of the ENHS thermal hydraulic optimization code for recent ENHS design improvements, UCBNE Internal Report, December 2005, 115 pages [15] Sienicki, J.J., Updated thermal hydraulic analyses for recent ENHS design improvements, Proc. of the 2003 International Congress on Advances in Nuclear Power Plants, Cordoba, Spain, May 4-7, 2003, paper No. 3115T [16] Okawa, T. and Greenspan, E. Effect of fuel type on the attainable power of the Encapsulated Nuclear Heat source Reactor, Proceedings of ICAPP-2006, Reno, NV, June 4-7, 2006