SMALL MODULAR REACTORS

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1 SMALL MODULAR REACTORS James T. (Tom) Voss, NRRPT, CHP Los Alamos National Laboratory Health Physics Measurements Group (RP-SVS) Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 1

2 Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 2

3 The defines small (Small Modular Reactor) as less than 300 MWe. Many hundreds of small nuclear reactors have been built for naval use up to 190 MWt. Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 3

4 In the US there are a few designs under serious consideration. In January 2012 the DOE called for applications from industry to support the development of one or two US light-water reactor designs, allocating $452 million over five years. Four applications were made, from Westinghouse, Babcock & Wilcox, Holtec, and NuScale Power, the units ranging from 225 down to 45 MWe. Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 4

5 In March 2013 the DOE called for applications for second-round funding, and proposals were made by Westinghouse, Holtec, NuScale, General Atomics, and Hybrid Power Technologies, the last two being for EM2 and Hybrid SMR, not PWRs. Other (non- PWR) small reactor designs will have modest support through the Reactor Concepts RD&D program. Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 5

6 A late application from left field was from National Project Management Corporation (NPMC) which includes a cluster of regional partners in the state of New York, South Africa s PBMR company, and National Grid, the UK-based grid operator with 3.3 million customers in New York, Massachusetts and Rhode Island. The project is for a HTR of 165 MWe, apparently the earlier direct-cycle version of the shelved PBMR Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 6

7 Applications to the NRC for construction of SMRs are expected in 2015 (or 2016). Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 7

8 A recent report states there are 131 SMRs currently operating around the world. Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 8

9 One of the big drivers for the Implementation of SMRs in the US is the Initial capital investment. Also, the environmental impact of a single SMR is less than for a traditional power reactor in the 1,000 MWe range. Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 9

10 At a recent HPS meeting an NRC representative stated that until the NRC received an application for an SMR they could not determine the security requirements. Then those security requirements might need to be codified before a license could be issued. Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 10

11 The number of personnel needed to operate an SMR will be much less than for a traditional power reactor. However, the NUMBER OF PERSONNEL PER MW OUTPUT WILL BE MUCH GREATER. This also applies to the number of radiation instruments required. Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 11

12 Questions? Comments? James Tom Voss, NRRPT, CHP Operated by Los Alamos National Security, LLC for the U.S. Department of Energy s NNSA U N C L A S S I F I E D Slide 12

13 LA-UR Approved for public release; distribution is unlimited. Title: Experimental Demonstration of a Heat Pipe/Stirling Engine Nuclear Reactor Author(s): Poston, David I. McClure, Patrick R. Dixon, David D. Gibson, Marc A. Mason, Lee S. Intended for: Nuclear Technology Issued: Disclaimer: Los Alamos National Laboratory, an affirmative action/equal opportunity employer,is operated by the Los Alamos National Security, LLC for the National NuclearSecurity Administration of the U.S. Department of Energy under contract DE-AC52-06NA By approving this article, the publisher recognizes that the U.S. Government retains nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or to allow others to do so, for U.S. Government purposes. Los Alamos National Laboratory requests that the publisher identify this article as work performed under the auspices of the U.S. Departmentof Energy. Los Alamos National Laboratory strongly supports academic freedom and a researcher's right to publish; as an institution, however, the Laboratory does not endorse the viewpoint of a publication or guarantee its technical correctness.

14 TBD Experimental Demonstration of a Heat Pipe/Stirling Engine Nuclear Reactor David I. Poston, Patrick R. McClure, and David D. Dixon Los Alamos National Laboratory MS C921, Los Alamos, New Mexico (505) : poston@lanl.gov Marc A. Gibson and Lee S. Mason National Aeronautics and Space Administration Glenn Research Center, Cleveland, Ohio, Abstract Los Alamos National Laboratory and Glenn Research with the help of National Security Technologies demonstrated the use of a nuclear fission system as a power source that transferred heat via a water based heat pipe to a small Stirling engine based power converter to produce electricity. This experimental setup demonstrated that a small reactor based on heat pipes and Stirling Engines is possible and produces a system with well-characterized nuclear feedback between the reactor and the power conversion system. This paper describes the experimental setup, modeling of the system, and results that confirm the basic physics of the experiment. Keywords: Heat pipe, Stirling engine, Reactor physics I. INTRODUCTION Space fission power has long been recognized as a technology that could enable truly ambitious space science and exploration; however, the US has not successfully developed, or even tested a space reactor concept for over 40 years. In general, past space fission power programs have failed because they have tried to do too much, too soon; thus they required lengthy and costly development. If a fission system is to be utilized in the near term, a reasonable, measured first step must be taken. Los Alamos National Laboratory (LANL) and Glenn Research Center (GRC) have recently proposed a very small, simple reactor that could provide ~1 kwe of electricity for outer-solar system space exploration. The proposed concept uses heat pipes to transfer power from a small uranium core to a Stirling-engine power conversion system. The small reactor design uses existing technology and lends itself to quick and affordable development. LANL and NASA have designed several nuclear fission reactors for space applications that are based on the physics of a fast reactor with heat pipes to transfer heat from the nuclear fuel to a Stirling engines as a means of power conversion[1-5]. The configuration is a solid block core with heat pipes in the solid core that extend up to the power conversion system. The proposed small 1 kwe reactor concept is shown in Figure 1. Fig. 1. Artist conception of the heat pipe/stirling Engine very small space reactor Heat pipe reactors have been proposed because of their inherent simplicity; the reactor is nearly solid state. The simplicity of the reactor concept leads to a reliable, safe, easy to operate and easy to manufacture reactor. One of the key features of these types of reactors is the simple and predictable reactivity feedback mechanism that allows for the reactor to be load following. Fast reactors in this size range are controlled by thermal expansion and subsequent negative reactivity feedback. Thermal feedback will lower the reactor power if less heat is extracted by the power conversion system. This makes the system more tolerant of power conversion failures. It also allows for above-rated power extraction if needed (within thermal limits of fuel system.) This phenomenon can be modeled, but a nuclear demonstration was needed to validate the modeling. More importantly, a demonstration was needed because a heat pipe cooled reactor had never previously been tested. The Demonstration Using Flattop Fission

15 TBD (DUFF) experiment was devised as a proof-of-concept test for the aforementioned small space reactor. The test was conducted in September of 2012 and was successful on many levels, but most importantly it produced electricity using nuclear fission, heat pipes, and Stirling engines. DUFF demonstrated the first Stirling engine to be powered with fission energy, the first use of a heat pipe to transport heat from a reactor to a power conversion system, and the first nuclear-powered test of a space reactor technology in over 40 years. One of the biggest benefits of the testing was that it provided valuable experimental data. The data demonstrated that the concept is viable and the physics is well characterized. The data also provided the ability to benchmark criticality and dynamic reactor modeling tools used to design the experiments. II. DUFF EXPERMIMENTAL SETUP The experiment to model the reactor consists of three main parts combined into a single experimental setup. The three parts of the experiment are: 1. The critical experiment assembly Flattop 2. A water based heat pipe 3. A Stirling Engine power conversion system The three pieces of the experiment were coupled together to form a complete nuclear system that could go critical, develop fission heat, transfer the heat via the heat pipe to the Stirling engine, and the Stirling engine producing electricity. Each piece of the experiment is described, as is the modeling of the system, a description of the test conditions, the results of the test and it s comparison to the modeling. II.A Flattop Critical Experiment Assembly The Flattop Critical Experiment Assembly was designed to provide benchmark neutronic measurements in spherical geometry, with a number of different fissile driver materials. Flattop includes a spherical core of highly enriched uranium (HEU), a natural uranium reflector, support structure, hydraulic system, and instrumentation and control systems. Flattop is shown in Figure 2. A distinguishing feature of Flattop that made the assembly ideal for this experiment was that the HEU spherical core and the natural uranium reflector contain a inch hole through the reflector and the core. This allows for a 0.5 inch heat pipe to placed into the core with no modification to the existing critical assembly. Fig. 2. Flattop critical experiment assembly with core The uranium core is comprised of a 4.77-inch diameter sphere weighing 16.2 kg of 93.15% U-235; additional HEU pieces bring the total core mass to ~18 kg. The uranium sphere rests on its own natural uranium pedestal. The reflector assembly is a 19-inch sphere of natural uranium mounted on a steel table supported by a floormounted base. The reflector assembly consists of two retractable quarter-spheres (~250 kg each) and a fixed hemisphere (~500 kg). II.B. Water Heat Pipe Flattop is a national asset, so the temperature of the experiment was limited to 300 C, not the 800 C of a possible flight concept. This required the use of water heat pipes instead of sodium, a chiller to provide cold-end temperatures below room temperature, and off the shelf (literally) Stirling convertors that could operate at these temperatures. The water heat pipe that connected Flattop and the Stirling engine converters utilizes a stainless-steel wall and a sintered-nickel wick. It has an outer diameter of approximately 0.5-inches and is approximately 45 inches in length. The heat pipes was developed and tested coupled to the Stirling Engine convertors at GRC prior to shipment to Nevada.

16 TBD II.C. Stirling Engines A free-piston Stirling engine with an integral linear alternator that converts the piston reciprocating motion to electrical output was used for the experiment. Two of these engines were used and connected to each other and the heat pipe by means of a cooper block. These Stirling engines were provided by Sun Power Inc. and GRC. They were specifically designed for, and tested at, low temperatures. With the other convertors, the springs will stiffen at low temps so for these a displacer mass was added to compensate; they can operate at -30 C on the cold end and up to 400 C on the hot end. They also provide positive power at 150 C, which is lower than any of the other convertors (which need 200 C). The Stirling engines are mechanically simple very similar to Stirling engines already in used by NASA for radioisotope power systems. The Stirling engines are shown in Figure 3. Fig. 4. Compete experimental assembly of DUFF experiment III. MODELING OF EXPERIMENT The modeling of the experiment consisted of two pieces: A detail Monte-Carlo neutronics model A dynamic system model with coupled heat transfer and neutronics. Each is described in more detail below. III.A. Detailed Monte Carlo Neutronics Model Fig. 3. Two Stirling engines with linear alternators connected by a copper block with a ½ hole for the heat pipe. II.D. Final Experimental Assembly The experiment was performed at the Nevada Nuclear Security Site (NNSS) Device Assembly Facility (DAF). The DAF is a Hazard Category 2, Security Category 1 nuclear facility owned by the National Nuclear Security Administration (NNSA). DAF is operated by National Security Technologies, LLC for the NNSA. The complete experimental setup is presented in Figure 4. A detailed model of the Flattop critical experiment was done with MCNP66. MCNP6 is a LANL Monte Carlo code for radiation transport. The MCNP code is widely used in studies of advanced reactor concepts, either directly as a main-line design tool or indirectly as part of the verification/validation process. MCNP is routinely used to calculate keff and detailed distributions of power and reaction rates. MCNP provides highly accurate results, using continuous-energy physics, ENDF/B-VII nuclear data, and explicit 3D constructive solid geometry The model was used to generate time dependent reactivity coefficients for the system s model. An illustration of the MCNP model of Flattop is shown in Figure 5.

17 TBD Fig. 5. MCNP model of the Flattop Critical Experiment with a water based heat-pipe. Neutronic calculations of the DUFF experiment were performed with MCNP6 using the ENDF/B-VII crosssection data. All MCNP cases were run with one billion source particles, resulting in a statistical error of Cross sections, including temperature dependent values for reactivity coefficient calculations, were generated by NJOY6 7. Criticality results were expressed in excess reactivity (the amount over critical) in the units of cents. The value of excess reactivity was based on measurement of the reactor period in a zero-power state, and calculated via the inhour equation 8. The conversion to/from cents was based on the MCNP calculated beta-effective, i.e. the neutronic worth of the delayed neutrons. The calculated betaeffective of is higher than the delayed neutron fraction of 235 U because a substantial number of fissions occur in the reflector 238 U (note: the difference in energy between fission and delayed neutrons does not have much impact on beta-effective in this reactor). The DUFF MCNP model was created using the original design drawings of the Flattop assembly. A core close up is shown in Fig. 6. The only difference between the model and the DUFF experiment is that the angular position of the HEU mass adjuster buttons (the stubs on the bottom) are different. This was done to simplify the model, but this approximation has a minimal impact because the radial position and the approximate spacing was preserved. The only part of the geometry that is not known in detail is the amount water in the heat pipe evaporator (shown by the white cylindrical region in the core center in Fig. 6). Fig. 6. MCNP Model of Flattop/DUFF Core. III.B. Systems Model The system model is a point kinetics neutronics model coupled to a two-dimensional heat transfer model. The model is based upon a Fortran based code developed at LANL called FRINK 9. FRINK uses a simple point kinetics solver coupled with finite-difference thermal calculations. The combined set of differential equations, estimated by the Crank-Nicolson method, is reduced with a matrix solver. FRINK models reactivity coefficients individually by component, which is very important for the Flattop critical experiment because the reflector heats slowly as compared to core. FRINK allows modeling of geometrically delayed neutron groups, which can be important in systems where a large fraction of fissions result from reflected neutrons. The heat transfer model includes all significant conduction and radiation paths in the system. This includes the modeling of heat transfer across several gaps between components: fuel-sphere-to-fuel-adapter, fuel-adapter-toheat-pipe, heat-pipe-to-reflector-adapter, reflector-adapterto-reflector, and heat-pipe-to-copper-block. Explicit models of each gap were added to FRINK, which included the option of filling each individual gap with air (plus thermal radiation) or a conducting medium. The model was modified to thermally expand all of the pieces, such that the gap size was varied as a function of temperature. The model includes temperature dependent thermal properties for each material. The model also includes a heat pipe model (including simplified heat pipe physics), a copper block to transfer heat to the Stirling engines and a representation of the energy removed by the operation of the Stirling engine.

18 TBD III.C. Pre-DUFF Model Benchmarking Soon after the DUFF experiment was proposed, the systems model was developed and benchmarked against a series of test runs on Flattop in These tests were a series of free runs done with excess reactivity on Flattop of 10, 20 and 30 cents each by Goda 10. These benchmarks provided confidence that the systems model would accurately reproduce the planned experimental DUFF runs with Flattop (i.e. the heat pipe and Stirling Engine configurations.) An example of the benchmarking is shown in Figure 7. Figure 7 shows a comparison of the calculated temperatures in Flattop during an excess reactivity run of 20 cents versus the actual temperature data. The results of the system model compare very well with the actual data. measured excess reactivity of $0.50.) The MCNP calculated value of k eff of the room temperature DUFF configuration was This is remarkably good agreement. There s no guarantee that the water was exactly as modeled, or that an unknown difference between the model and experiment existed, but the most likely scenario is that this is indeed an excellent benchmark. The second zero-power critical used the actual DUFF heat pipe, which was overfilled with water because it was found to boost performance during testing at GRC. For this configuration the experimental value of k eff was determined to be (based on an excess reactivity of $0.67). Therefore, the extra water in the pipe had added $0.17 of reactivity. MCNP calculated the worth of fully flooding the evaporator region as $0.41, so this indicates that a significant fraction (~40%) of the evaporator vapor region was filled with water before the heat pipe began to operate. This had to be taken into account in the dynamic modeling, because as soon as the evaporator heated up the pool evaporated and/or was pushed up by pressure. It was no surprise that an MCNP calculation of a fastspectrum uranium based system provided an accurate value of k eff. In this case a very close comparison could be expected because every part of the Flattop assembly was modeled based on original design drawings, plus 235 U is the most studied cross section of any isotope, so uncertainty in the model and data was low. Fig. 7. Benchmark of FRINK system model temperatures to actual Flattop temperature data for a 20 cent excess reactivity run. III.D. Predicted Behavior as Basis for Experiment The systems model was then used to predict the behavior of the system when the heat pipe and Stirling engines were added to the system. These calculations provided a basis for writing the experimental plan for the planned tests on the actual nuclear system. Tests were planned based on the rate of reactivity insertion for Flattop. Ideally several tests were planned for execution, but time and funding restrictions allowed for only two tests to be performed, a fast insertion of excess reactivity and a slow insertion of excess reactivity. Both experiments were designed to use up to the maximum amount of excess reactivity allowed by the regulatory limits of the Flattop critical experiment machine of 80 cents. The first zero-power approach to critical test used the dummy heat pipe, which had the same water loading as was used in the model. The room temperature experimental value of k eff for this configuration was (based on a The DUFF experiment introduced stainless steel (heat pipe wall), nickel (wick), and water into the most reactive region of the reactor. This introduced absorber and moderator in the core, and a complex mix of reflection and absorption in the reflector region. The worth of the heat pipe was calculated as only $0.18 (positive) but only because the positive impact of reflection and moderation was mostly offset by parasitic absorption. The highly accurate calculation of k eff (as compared to the experimental value) is more impressive with these materials included. IV. EXPERIMENTAL RESULTS The experiment was performed twice, once on September 13 th, 2012 and the second time on September 18 th, The first experiment was a fast ramp up on reactivity insertion for the experiment. The second experiment was a slower ramp up on reactivity insertion. Each is described in detail below. IV.A. Fast Reactivity Insertion Results The timeline of events for this experiment are provided below in minutes and correspond the set of events shown graphically in Figure 8.

19 TBD T=0 min: Reactor operator starts to take power up to moderate level (~2 kw t ), a small amount of heating occurs. T=5 min: Operator inserts reactivity (~30 cents) and power quickly rises to about ~10 kw t. Natural temperature feedback lowers reactivity, and operator continually inserts reactivity via control rods to maintain power for ~1 minute. T=5+ min: The model indicates a fuel temperature rise of ~2 C/sec, reaching 300 C in just a few minutes. T=6 min: Despite the hot core, the cold end of the Stirling engine remains below room temperature (due to chiller on cold end of Stirling). System waits for the water in the heat pipe to establish its internal 2-phase passive circulation loop. T=7 min: The heat pipe turns on and heat flows to the Stirlings. The heat pipe power quickly goes to about 400 W t. T=7+ min: Heat pipe power causes a quick rise in the heat pipe temperature just past evaporator, and then begins to heat copper block and Stirling engine hot side. T=8 to 17 min: As the heat pipe draws power from the reactor, it heats up the Stirling by about 200 C over ~10 minutes, while also keeping the core cooler (thus adding more reactivity and causing additional reactor power). T=17min: Stirling engine technician starts the engine when the hot end is 225 C, providing 24 W e of electricity. T=17+ min: The Stirling hot end cools quickly (thus sharp drop in electrical output). Temperature gradients start to set up to allow steady-state power flow from core over the next minute. Stirling power starts to level off at ~18 W e. T=18 min: Reactor scrammed (fission power to ~zero). T=18+ min: Stirling draws power from stored energy and decay heat for ~8 minutes, cooling all components by ~100 C. Stirling continues to produce electricity, but at diminishing efficiency and power as temperature drops. T=26 min: Stirling engine stalls when temperature hits ~120 C. The test results showed the expected negative temperature feedback behavior of the reactor physics. Fast reactors in this size range are controlled by thermal expansion and subsequent negative reactivity feedback. The simplest representation of this phenomenon is in the fuel. If the fuel heats up then the atoms expand further apart, therefore it is easier for more neutrons to leak out of the fuel and avoid a fission reaction. This leakage results in lower reactivity (lower neutron multiplication), and thus decreases the slope of power production. If the system is operating at steady state, the slope of power will go from zero to negative, thus system power decreases. Therefore, during nominal operation, if more heat is extracted by the power conversion system then thermal feedback will raise reactor power, or vice-versa. In this manner, the reactor is a load following system component; i.e. the reactor thermal power automatically matches itself to the thermal power removed by the Stirling engines. There are some 2 nd order effects that complicate this scenario, but for compact, fast reactors they are relatively small. The system also showed that a heat piped cooled reactor coupled to a Stirling Engine power conversion was possible and practical. This was the first nuclear demonstration of a heat piped cooled reactor. IV.B. Slow Reactivity Insertion Results The timeline of events for this experiment are provided below in minutes and correspond the set of events shown graphically in Figure 9. This test was a repeat of the September 13 th, 2012 test but with a different rate of reactivity insertion. T=0 min: Reactor operator inserts reactivity and slowly increases power. T=2+ min: The model indicates a fuel temperature rise of ~15 C/min (0.25 C/sec) over next ~7 minutes. T=8 min: Reactor operator levels off power at ~2 kw t, holds power with small reactivity insertions over next 8 minutes. T=9 min: Heat pipe turns on and starts cooling core and heating Stirling hot end. T=16 min: Operator begins series of insertions to increase fission power from 2 to 3.5 kw t over next ~20 min. T=22 min: The temperature of heat pipe and the Stirlings converge; i.e. heat flux from the heat pipe has fully soaked the thermal inertia of the copper block and Stirling all components at ~160 C. T=22+ min: Over the next 8 minutes, the core temperature continues to increase due to fission power. The heat pipe and Stirling track the core temperature increase in tandem, at rate of ~5 C/min. Fig. 8. Data from September 13 th run of DUFF Experiment.

20 TBD T=30 min: Stirling engine technician starts the engine at a reduced engine stroke (a lower power setting) at a hot head temperature of 180 C, the electric power output is 13 We. T=31 min: The power draw from the engine quickly cools the hot head of the Stirling and the power level drops to a steady state value of ~7 We. T=32 min: Over the next 8 minutes electric power tracks increases in power and temperature, from 7 We to 11 We. T=35 min: Reactor operator inserts a large final reactivity insertion, bringing peak fission power to ~5.5 kwt. T=36+ min: Reactor power quickly drops due to negative temperature feedback, from 5.5 kwt to ~3 kwt in 5 minutes. T=39 min: Stirling technician increases stroke of the engine causing electric output to increase from 11 We to 17 W e. T=42 min: Stirling technician stops engine. Engine hot end temperature increases from 185 C to 225 C. T=44 min: Stirling technician restarts engine. Reconfirms electric power of 24 We at same 225 C as September 13th run. T=46 min: Reactor scrammed (fission power to ~zero). T=46+ min: Stirling draws power from stored energy and decay heat for ~5 minutes. T=54 min: Stirling engine stalls when temperature hits ~115 C. The systems model described in Section III.B was used to compare the data collected to the model results. Note the rate at which the reactor operator inserted reactivity was unknown until after the experiment. The actual reactivity insertions of the experiment were used as input to the systems model and the results of the neutronic and thermal models then compared to the actual data. For the September 13th runs the comparison of the results is shown in Figure 10. Solid lines are data collected from the experiment. Dashed lines are the system model results. Fig. 10. Comparison of experimental data to model results from September 13th run, temperatures in degree C. Fig. 9. Data from September 18th run of DUFF Experiment. The test again behaved as expected with reactor power changing to accommodate changes in the power conversion system (see events at about 36 minutes in test.) This result validated the expected changes given the negative temperature feedback of the reactor. IV.C. Comparison of Data to Systems Model The model compares very well with the experimental data for the heat pipes and Stirling engines. The predicted peak temperature of the reflector does not match well with the reflector pedestal largely because they are measuring the temperature at different locations. The good agreement between predicted and actual values for DUFF was not an unexpected result. It is generally accepted that the point kinetic solution of neutron population is valid for compact, fast-spectrum reactors, but it was still very encouraging to see the FRINK results match the experimental results. The well-behaved system was assumed to be very easy to model and this result was confirmed. There were certain aspects of the FRINK solution that did not ideally match the experiment, but those were all in regions were the temperatures were rapidly changing (such that the thermal approximations were likely inaccurate). The September 18th run the comparison is shown in Figure 11.

21 TBD applications were low-power, reliability and simplicity are key requirements. VI. NOMENCLATURE Fig.11. Comparison of experimental data to model results from September 18 th run, temperatures in Degree C. The experimental results once again compared very well to model simulation. This provides confidence that a fast reactor system with heat pipes and Stirling engines is a very simple system that is easy to model and characterize. As the results show, the calculated reactivity temperature coefficients (RTCs) were able to reproduce the experimental data very well. The RTCs were calculated based on elevated temperature MCNP calculations, In addition, the system transient calculations used an additional RTC based on the spreading of the reflector segments caused by core expansion. The reflector movements create very thin gaps between the segments, which were modeled by MCNP to determine their impact on reactivity. These reflector RTCs, which become substantial when the fuel contacts the reflector, did a nice job of recreating the experimental data. V. SUMMARY LANL and Glenn Research with the help of National Security Technologies have demonstrated that a heat pipe cooled fast reactor with Stirling Engine power conversion is a practical reactor concept and would be a very good design for space reactor applications. This test was the first demonstration of a heat pipe cooled reactor of any size. Heat pipe cooled reactors have been postulated in the past and tested using separate effects testing (i.e. electrically heated), but this is the first heat pipe reactor using nuclear fission as a heat source. The test demonstrated that a heat pipe cooled fast reactor with Stirling engine power conversion has well characterized physics that can be easily modeled. This test will serve as the basis for future investigations into heat pipe cooled/stirling engine reactors for space reactor LANL Los Alamos National Laboratory GRC Glenn Research Center DUFF Demonstration Using Flattop Fissions We Watt electric Wt Watt thermal HEU Highly Enriched Uranium MCNP Monte Carlo N-Particle RTC Reactivity temperature coefficients NNSS Nevada National Security Site DAF Device Assembly Facility NNSA National Nuclear Security Administration VII. REFERENCES 1. D. R. Koenig, W. A. Ranken, Heat Pipe Nuclear Reactors for Space Applications, Los Alamos Scientific Laboratory report LA-UR (1977). 2. W. A. Ranken, M. G. Houts Heat Pipe Cooled Reactors for Multi-Kilowatt Space Power Supplies, Los Alamos National Laboratory report LA-UR (1994). 3. M. G. Houts, D. I. Poston, W. A. Ranken, Heat Pipe Space Power and Propulsion Systems, Los Alamos National Laboratory report LA-UR (1995). 4. D. I. Poston, Nuclear Design of the SAFE-400 Space Fission Reactor, Nuclear News (2002). 5. D. I. Poston, Prometheus 1 Design Studies Report, Los Alamos National Laboratory report LA-CP (2005). 6. D. B. Pelowitz, MCNP6 User's Manual, Los Alamos National Laboratory report LA-CP (2011). 7. R. E. McFarlane, R. J. Barrett, D. W. Muir, and R. M. Boicourt, NJOY Nuclear Data Processing System. User's Manual, LA M (1978) 8. Keepin G.R. Physics of Nuclear Reactor Kinetics, Addison-Wesley, Reading, Mass., (1965) 9. D. I. Poston, D.I., Dixon, D.D, Marcille, T.F., Amiri, B.W., FRINK A Code to Evaluate Small Reactor Transients, Proceedings of Space Technology and Applications International Forum (STAIF-2007), AIP Conf. Proc. #880, Melville, NY (2007).

22 10. J. M. Goda, J. A. Bounds, W. L. Myers, and R. G. Sanchez, Flattop Free-Run Demonstration, Los Alamos National Laboratory report LA-UR (2012). TBD

23 Presented at the 24 th Meeting of the TWG - NPPIC, on 23 May 2013 at s Press Room Development Status of Small and Modular Reactors M. Hadid Subki Technical Lead for SMR Technology Development Nuclear Power Technology Development Section Division of Nuclear Power, Department of Nuclear Energy International Atomic Energy Agency

24 Outline What s new in global SMR development and deployment? Reactors Under Construction in SMR category Practical Categorization of SMR Design & Technology Perceived Advantages and Challenges Anticipated Issues on I&C Project on SMR Technology Development Events on SMR in P&Bs and Summary 2

25 What s New in Global SMR Development? mpower NuScale W-SMR Hi-SMUR SMART KLT-40s SVBR-100 BREST-300 SHELF Flexblue B&W received US-DOE funding for mpower design. The total funding is 452M$/5 years for 2 out of 4 competing ipwr based-smrs. Some have utilities to deploy in specific sites. US-DOE also announced the second round of SMR funding in March On 4 July 2012, the Korean Nuclear Safety and Security Commission issued the Standard Design Approval for the 100 MWe SMART the first ipwr received certification. Construction of 2 modules of barge-mounted KLT-40s near completion; Lead Bismuth cooled SVBR-100 & Lead-cooled BREST-300 to deploy by 2018, SHELF seabed-based conceptual design DCNS originated Flexblue capsule, 160 MWe, m seabed-moored, 5-15 km from the coast, off-shore and local control rooms CAREM-25 Site excavation for CAREM-25 was started in September 2011; construction started in S Toshiba had promoted the 4S for a design certification with the US NRC for application in Alaska and newcomer countries. PFBR PHWRs: 220, 540 & 700, AHWR300-LEU The Prototype FBR ready for commissioning and start-up test. 4 units of PHWR-700 under construction, 4 more units to follow. AHWR300-LEU at final detailed design stage and ready for construction. 3

26 What s New in Global SMR Development? (cont d) CEFR HTR-PM ACP-100 IRIS 2 modules of HTR-PM under construction; CNNC developing ACP-100 which will be deployed by 2018 Politecnico di Milano (POLIMI) and universities in Croatia & Japan are continuing the development of IRIS design - previously lead by the Westinghouse Consortium Recently introduced at the 2012 SMR Meetings: ACP-100, CNNC, China Flexblue, DCNS, France 4

27 Reactors Under Construction in SMR category Country Argentina China India Reactor Model CAREM-25 (a prototype) HTR-PM (GCR) (a prototype) Output (MWe) Designer Number of units Site, Plant ID, and unit # Commercial Start 27 CNEA 1 CAREM ~ Tsinghua Univ./Harbin 1 (2 modules) Shidaowan unit ~ 2018 PHWR NPCIL 2 Kakrapar 3 and 4 6/2015 and 12/2015 PHWR NPCIL 2 Rajashtan units 7 and 8 6/2016 and 12/2016 PFBR 500 (LMFBR) Pakistan CNP CNNC - China Russian Federation KLT-40S (ship-borne) 500 IGCAR 1 PFBR Kalpakkam OKBM Afrikantov 2 Chasnupp 3 and 4 12/ Akademik Lomonosov

28 Practical Categorization of SMRs Advanced SMRs including modular reactors and integrated PWRs Innovative SMRs including those of Gen-IV reactors with non-water coolant/moderator Converted and Modified SMRs Including barge-mounted floating NPP and seabed-moored submarine-like reactors Conventional SMRs Those of 1970/80s technologies and still being deployed Innovative Application of SMRs with Non-Nuclear Including Nuclear-Renewable Hybrid System, and SMRs coupled with Non-Electric Applications (Desalination, H 2 production) 6

29 Practical Categorization of SMRs Advanced SMRs (incl. Modular and integrated-pwrs) CAREM-25 Argentina SMART Korea, Republic of VBER-300 Russia WWER-300 Russia ABV-6 Russia HTR-PM China mpower USA NuScale USA Westinghouse SMR - USA CEFR China 4S Japan PFBR-500 India 7

30 Concept of Integral PWR based SMRs SMART Westinghouse SMR pressurizer CRDM pumps Steam generators Steam generators CRDM core + vessel pumps core + vessel 8

31 Integral Primary System Configuration Courtesy: Westinghouse Electric Company LLC, All Rights Reserved X X X X X X XX X Benefits of integral vessel configuration: eliminates loop piping and external components, thus enabling compact containment and plant size reduced cost Eliminates large break loss of coolant accident (improved safety) 9

32 (cont d) Practical Categorization of SMRs Advanced SMRs (incl. Modular and integrated-pwrs) Each module has a dedicated turbine generator Modularity permits scaling to any size Generator Steam Turbine Condenser Water-Filled Pool Below Ground Containment NSSS J. Nylander and M. Cohen Courtesy of NuScale Power, USA. 10

33 (cont d) Practical Categorization of SMRs Innovative SMRs IMR Japan AHWR300-LEU India GT-MHR USA PRISM USA EM 2 USA PBMR South Africa 11

34 (cont d) Practical Categorization of SMRs Converted/Modified SMRs KLT-40s Russian Federation SVBR-100 Russian Federation Flexblue France 12

35 (cont d) Practical Categorization of SMRs Conventional SMRs VVER-440, Russian Federation 13

36 (cont d) Practical Categorization of SMRs Innovative Application of SMRs with Non-Nuclear Composite Wind Farms Variable Electricity Max Output of 1061 MWe to the power GRID Node Regional Biomass (80 Km radius or ~2 million hectares) Reference: a study by the EC JRC t/dm/yr 1018 MWe Nuclear reactor 347 MWe (755 MWth) Offsetting SMR Electricity Reactor Heat Dynamic Energy Switching Hydrogen Electrolysis 104 GWh heat at 200 C 1169 GWh heat at 500 C t H 2 /yr Drying and Torrefaction Processes + Pyrolysis +Synfuel Production Torrified Product Pyrolyzed oil + char + offgas 753m 3 /day bio-diesel 597m 3 /day bio-gasoline 14

37 (cont d) Practical Categorization of SMRs Innovative Application of SMRs with Non-Nuclear Reference: a study by the US DOE 15

38 Perceived Advantages and Challenges Observation Technological Issues Non-Technological Issues Advantages Shorter construction period (modularization) Potential for enhanced safety and reliability Design simplicity Suitability for non-electric application (desalination, etc.). Replacement for aging fossil plants, reducing GHG emissions Fitness for smaller electricity grids Options to match demand growth by incremental capacity increase Site flexibility Reduced emergency planning zone Lower upfront capital cost (better affordability) Easier financing scheme Challenges Licensability (due to innovative or first-of-a-kind engineering structure, systems and components) Non-LWR technologies Operability performance/record Human factor engineering; operator staffing for multiple-modules plant Post Fukushima action items on design and safety Economic competitiveness First of a kind cost estimate Regulatory infrastructure (in both expanding and newcomer countries) Availability of design for newcomers Infrastructure requirements Post Fukushima action items on institutional issues and public acceptance 16

39 Anticipated Issues on I&C Specific Operational and Process Characteristics First of a kind structure, system and component engineering; New plant dynamic behaviour and special architectures Innovative reactors with long fuel cycles and extended operation Non-water coolants/moderators and extreme remote environments Improved Functionality Multi-modules plant operation Flexible operation (requirements for extreme load-follow capability) Non-electric applications Control of Nuclear-RES Hybrid energy systems I&C to satisfy specific safety requirements in post-fukushima Revisit defence-in-depth, diversity, redundancy, independency of safe reactor trip versus ESF actuations I&C for Non-Electric Emergency Core Cooling Systems Technological Needs Sensor technologies for integral-pwrs process monitoring and measurements Digital I&C Cyber-Security & Communications SMRs deployment in remote-areas Emergency planning zone Reduced Staffing and so forth.. to be discussed. 17

40 SMR for Immediate Deployment CAREM-25 Full name: Central Argentina de Elementos Modulares Designer: National Atomic Energy Commission of Argentina (CNEA) Reactor type: Integral PWR Coolant/Moderator : Light Water Neutron Spectrum: Thermal Neutrons Thermal/Electrical Capacity: 87.0 MW(t) / 27 MW(e) Fuel Cycle: 14 months Salient Features: primary coolant system within the RPV, self-pressurized and relying entirely on natural convection. Design status: Site excavation started for construction in 2012

41 SMR for Near-term Deployment 2011 KAERI Republic of Korea SMART Full name: System-Integrated Modular Advanced Reactor Designer: Korea Atomic Energy Research Institute (KAERI), Republic of Korea Reactor type: Integral PWR Coolant/Moderator: Light Water Neutron Spectrum: Thermal Neutrons Thermal/Electrical Capacity: 330 MW(t) / 100 MW(e) Fuel Cycle: 36 months Salient Features: Passive decay heat removal system in the secondary side; horizontally mounted RCPs; intended for sea water desalination and electricity supply in newcomer countries with small grid Design status: Standard Design Approval just granted on 4 July 2012

42 SMR for Near-term Deployment: mpower Full name: mpower Designer: Babcock & Wilcox Modular Nuclear Energy, LLC(B&W), United States of America Reactor type: Integral Pressurized Water Reactor Coolant/Moderator: Light Water Neutron Spectrum: Thermal Neutrons Thermal/Electrical Capacity: 530 MW(t) / 180 MW(e) Fuel Cycle: 48-month or more Salient Features: integral NSSS, CRDM inside reactor vessel; Passive safety that does not require emergency diesel generator Design status: Design Certification application expected in 4 th Quarter of 2013

43 SMR for Near-term Deployment NuScale Full name: NuScale Designer: NuScale Power Inc., USA Reactor type: Integral Pressurized Water Reactor Coolant/Moderator: Light Water Neutron Spectrum: Thermal Neutrons Thermal/Electrical Capacity: 165 MW(t)/45 MW(e) Fuel Cycle: 24 months Salient Features: Natural circulation cooled; Decay heat removal using containment; built below ground Design status: Design Certification application expected in 4th Quarter of 2013

44 SMR for Near Term Deployment 2011 JSC AKME Engineering SVBR-100 Designer: JSC AKME Engineering Russian Federation Reactor type: Liquid metal cooled fast reactor Coolant/Moderator: Lead-bismuth System temperature: 500 o C Neutron Spectrum: Fast Neutrons Thermal/Electric capacity: 280 MW(t) / 101 MW(e) Fuel Cycle: 7 8 years Fuel enrichment: 16.3% Distinguishing Features: Closed nuclear fuel cycle with mixed oxide uranium plutonium fuel, operation in a fuel selfsufficient mode Design status: Detailed design

45 SMR for Near-term Deployment 2011 TOSHIBA CORPORATION 4S Steam Generator Reactor Turbine/ Generator Full name: Super-Safe, Small and Simple Designer: Toshiba Corporation, Japan Reactor type: Liquid Sodium cooled, Fast Reactor but not a breeder reactor Neutron Spectrum: Fast Neutrons Thermal/Electrical Capacity: 30 MW(t)/10 MW(e) Fuel Cycle: without on-site refueling with core lifetime ~30 years. Movable reflector surrounding core gradually moves, compensating burn-up reactivity loss over 30 years. Salient Features: power can be controlled by the water/steam system without affecting the core operation Design status: Detailed Design

46 TM on I&C for Advanced SMRs Being held this week: May at A 0742 Objectives: Discuss global status of SMR design development; Update on 's activities on I&C system engineering; Discuss advances in I&C architecture of several advanced SMRs under development for near term deployment; Identify issues on I&C arises from specific dynamic behaviour of SMRs; Develop outline of Technical Document and Document Preparation Proposal "Instrumentation and Control for Advanced SMRs for ; and Formulate follow-up activities in collaboration with NENP/NPES and related Departments/Divisions. 24

47 Roles of on SMR Development Facilitate efforts of Member States in identifying key enabling technologies in development and addressing key challenges in deployment; Establish and maintain international networks with Member States, industries, utilities, stakeholders; Ensure coordination of Member State experts by planning and implementing training programme and knowledge transfer through technical meetings and workshops Develop international recommendations and guidance focusing on specific needs of newcomer countries 25

48 Project on SMR Technology Development Project : Common Technologies and Issues for SMRs Objective: To facilitate the development of key enabling technologies and the resolution of enabling infrastructure issues common to future SMRs Activities ( ): Formulate roadmap for technology development incorporating safety lessons-learned from the Fukushima accident Review newcomer countries requirements, regulatory infrastructure and business issues Define operability-performance, maintainability and constructability indicators Develop guidance to facilitate countries with planning for SMRs technology implementation 26

49 Events on SMR in Technical Meeting on Instrumentation and Control for Advanced SMRs, in, Vienna, May; 2. INPRO Dialogue Forum on Licensing and Safety Issues of SMRs, in, Vienna, 29 July 2 August; 3. TM on SMRs Technology Development for Near Term Deployment, 2-4 September 2013, CNNC Chengdu, People Republic of China 4. TM on Environmental Impact Assessment for (SMRs) Deployment in Newcomer Countries, in, Vienna, October 5. Consultancy Meeting on Instrumentation and Control for Advanced SMRs, in, Vienna, November (please contact also Mr. Janos Eiler) 6. Workshop on Design Requirements for SMR and Advanced Reactor Technologies in Post-Fukushima Era, initially on December 2012 (EB Japan PUI) a postponed TM from

50 Thank you for your attention. For inquiries, please contact: Dr. M. Hadid Subki 28