Supercritical-Water Experimental Setup for Out-of-Pile Operation M. Miletić1, M. Růžičková1, R. Fukač1, I. Pioro2 and E. Saltanov2
1Research Centre Rez Ltd., Rez, Czech Republic, marija_miletic@live.com
2University of Ontario Institute of Technology, Oshawa, Canada, Igor.Pioro@uoit.ca
ABSTRACT
The main goal of the Generation-IV nuclear-energy systems is to address the fundamental research and development issues, which are necessary to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a SuperCritical Water-cooled Reactor (SCWR), which naturally continues utilization of well-known light-water-reactor technology.
Research Centre Rez Ltd. took part in a large European joint-research project dedicated to Generation-IV light-water reactors with motivation to contribute to the fundamental research and development of the SCWRs by designing and building a test facility called “SuperCritical Water Loop (SCWL)”. The loop’s main objective is to serve as an experimental facility for corrosion of materials for in-core as well as out-of-core structures, for testing and optimization of suitable water chemistry for future SCWRs and for studies of water radiolysis at supercritical conditions. This paper summarizes the concept of the SCWL in the Research Centre Rez Ltd., its design, utilization and first results obtained from non-active tests already performed within the supercritical-water conditions.
1. INTRODUCTION
Prior to a general discussion on SuperCritical Water-cooled Reactor (SCWR) concepts it is important to define special terms and expressions used at these conditions. For better understanding of these terms and expressions Fig. 1 is shown below.
1.1. Definitions of Selected Terms and Expressions Related to Critical and Supercritical Regions [1]
Compressed fluid is a fluid at a pressure above the critical pressure, but at a temperature below the critical temperature.
Critical point (also called a critical state) is a point in which the distinction between the liquid and gas (or vapour) phases disappears, i.e., both phases have the same temperature, pressure and volume or density. The critical point is characterized by the phase-state parameters Tcr, Pcr and vcr (ρcr).
Pseudocritical line is a line, which consists of pseudocritical points.
Pseudocritical point (characterized with Ppc and Tpc) is a point at a pressure above the critical pressure and at a temperature (Tpc > Tcr) corresponding to the maximum value of the specific heat at this particular pressure.
Supercritical fluid is a fluid at pressures and temperatures that are higher than the critical pressure and critical temperature. However, in the present paper, a term supercritical fluid includes both terms – a supercritical fluid and compressed fluid.
Supercritical “steam” is actually supercritical water, because at supercritical pressures fluid is considered as a single-phase substance. However, this term is widely (and incorrectly) used in the literature in relation to supercritical “steam” generators and turbines.
Superheated steam is a steam at pressures below the critical pressure, but at temperatures above the critical temperature.
Temperature, oC
200 250 300 350 400 450 500 550 600 650
Pressure, MPa
5.0 7.5 10.0 12.5 15.0 17.5 20.0 22.5 25.0 27.5 30.0 32.5 35.0
Critical Point
Pseudocritical Line
Liquid
Steam Saturation Line
Superheated Steam Supercritical Fluid
High Density (liquid-like)
Low Density (gas-like)
Tcr=373.95o C Pcr=22.064 MPa Compressed Fluid
Figure 1. Pressure-Temperature diagram for water.
1.2. General Considerations
The demand for clean, non-fossil-based electricity is growing; therefore, the world needs to develop new nuclear reactors with higher thermal efficiency in order to increase electricity generation and decrease detrimental effects on the environment. The current fleet of nuclear-power plant is classified as Generation II and III. However, all these designs (here we are talking about only water-cooled reactors) are not as energy efficient as they could be, because their operating temperatures are relatively low, i.e., below 350°C for a reactor coolant and even lower for a steam-power cycle.
The Generation-IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts. However, accounting that the vast majority of modern power nuclear reactors are water-cooled reactors we consider an SCWR concept as the most viable option for further development. The main objectives of using supercritical water in nuclear reactors are to: 1) Increase thermal efficiency of modern Nuclear Power Plants (NPPs) from 30 – 35% to about 45 – 50%, 2) Decrease capital and operational costs and hence, decrease electrical-energy costs, and 3) Allow for co-generation of hydrogen.
SCW NPPs will have much higher operating parameters compared to modern NPPs (a pressure of about 25 MPa and an outlet temperature up to 625ºC), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated.
Also, higher supercritical-water temperatures allow direct thermo-chemical production of hydrogen at low cost due to increased reaction rates [2].
The design of SCW nuclear reactors is seen as the natural and ultimate evolution of today’s conventional water-cooled reactors [1]. Development of SCWRs is based on the following three proven technologies: 1) modern PWRs, which operate at pressures of 15 – 16 MPa, i.e., rather high pressures; 2) BWRs, which have a once-through or direct-cycle design;
and 3) modern supercritical turbines with pressures about 23.5 - 38 MPa and inlet temperatures up to 625ºC, which operate successfully at coal-fired thermal-power plants for
more than 50 years. In addition, some experimental reactors used nuclear steam reheat with outlet steam temperatures well beyond the critical temperature (up to 550°C), but at pressures below the critical pressure (3 – 7 MPa), to increase the gross thermal efficiency of NPP (for details, see Fig. 2 and Tables 1 and 2 [3].
The SCWR concepts therefore follow two main types, the use of either: (a) a large reactor pressure vessel with a wall thickness of about 0.5 m to contain the fuelled reactor core, analogous to conventional PWRs and BWRs, or (b) distributed pressure tubes or channels analogous to conventional CANDU® nuclear reactors. In general, mainly thermal-spectrum SCWRs are currently under development worldwide.
1.3. Pressure-vessel SCWRs
The pressure-vessel SCWR concept is developed mainly in China, European Union (EU), Japan and some other countries. This type of reactor uses a traditional high-pressure circuit layout. However, due to significantly reduced flow rates, high outlet temperatures and some other parameters like significant fuel-sheath temperature non-uniformities, may appear, which in turn can lead to sheath damage. Another challenge associated with pressure-vessel SCWRs is the manufacturing pressure vessel due to quite large wall thickness. Also, in pressure-vessel reactors nuclear steam reheat at subcritical pressures is not possible. More information on thermal and fast pressure-vessel SCWRs can be found in the latest book [4].
Figure 2. Beloyarsk NPP (Russia) reactors schematic: (a) Unit 1 with indirect steam cycle and (b) Unit 2 with direct steam cycle (courtesy of Dr. Yurmanov, Russia):
– Reheated steam; – Saturated steam;
– Water-steam mixture; – Water.
Table 1. Main parameters of Beloyarsk NPP reactors [3].
Parameters Unit 1
(730 EChs & 268 SRChs)
Unit 2 (732 EChs & 266 SRChs)
Electrical power, MWel 100 200
No. of K-100-90-type turbines 1 2
Inlet-steam pressure, MPa 8.5 7.3
Inlet-steam temperature, ºC 500 501
Gross thermal efficiency, % 36.5 36.6
Uranium load, t 67 50
Uranium enrichment, % 1.8 3.0
Square lattice pitch, mm 200 200
Core dimensions, m: OD / Height 7.2 / 6 7.2 / 6
EChs – Evaporative Channels; SRChs - Steam-Reheat Channels
Table 2. Average parameters of Beloyarsk NPP Unit 1 before and after installation of SRChs [3].
Parameters Before SRChs
installation
After SRChs installation
Electrical power, MWel 60 – 70 100 – 105
Steam inlet pressure, MPa 5.9 – 6.3 7.8 – 8.3
Steam inlet temperature, ºC 395 – 405 490 – 505
Exhaust steam pressure, kPa 9 – 11 3.4 – 4.0
Water mass flow rate (1st loop), kg/h 1400 2300 – 2400
Gross thermal efficiency, % 29 – 32 35 – 36
Electrical power for internal needs, % 10 – 12 7 – 9 1.4. Pressure-channel SCWRs
The pressure-channel SCWR designs are developed mainly in Canada (Figure 3) and Russia [1]. Within those two main classes, pressure-channel reactors are more flexible to flow, flux and density changes than pressure-vessel
reactors. In addition, nuclear steam reheat can be implemented inside a pressure-channel SCWR based on the experience obtained during an operation of experimental BWRs in 60-s and 70-s, which makes it suitable to modern supercritical direct single- steam-reheat-cycle turbines. All these factors make it possible to use the experimentally confirmed, better solutions developed for these reactors. One of them is channel-specific flow-rate adjustments or regulations. Also, a pressure tube at such pressures will have a wall thickness of about 7 – 9 mm (see Fig. 4) compared to about 400 – 500 mm for a pressure vessel. Therefore, a design whose basic element is a channel, which carries a high pressure, has an inherent advantage of greater safety than large vessel structures at supercritical pressures.
Figure 3. Vertical core-configuration
option (courtesy of AECL)
Figure 4. High-Efficiency Channel (HEC) with ceramic insert (AECL design) and Re- Entrant Channel (REC) with annulus gas as thermal insulation for SCWR with liquid
moderator (drawings prepared by W. Peiman, UOIT).
However, the major problem for SCWRs development is reliability of materials at high pressures and temperatures, high neutron flux and aggressive medium such as supercritical water. Unfortunately, up till now nobody has tested candidate materials at such severe conditions (some candidate materials proposed by Russian scientists are listed in Table 3.
Therefore, testing candidate materials and water chemistry within operating conditions of SCWRs (pressures 24 – 26 MPa; temperatures 280 – 625°C and neutron flux about or higher than 1018 n/m2s) in an in-pile supercritical water loop is the most important task of today.
Table 3. Candidate materials for SCWR applications (based on Russian literature) [6].
Steel grade С Cr Ni Mo Nb N2 Ti Mn B V Ce Cu
Aust. st.st.: Fuel sheath; up to 650°С; up to 80–90 dpa
EI–847
09Сr16Ni15Mo3Nb <0.09 15–17 14–16 2.5–3.0 0.6–0.9
EP–172
0Сr16Ni15Mo3NbB 0.05–0.09 15–17 14–16 2.5–3.0 0.4–1.0 <0.05
ChS–68
0Сr17Ni14Mo2Ti 0.06 17 13.5 2–2.5 0.6–0.9 0.4 1.6 0.005
EI–844
03Сr16Ni15Mo3 <0.04 15–17 14–16 2.5–3.0
Fe–Ms st.: Bundle sheath; up to 600° С; up to 150–200 dpa
EP–450 1Сr12MoNbV 0.12 12 1.5 0.5 0.005
EI–756 11Сr11В2MoV
EP–450 –ODS
Powder technology Including nanotechnology with oxides Y2O3+0.5Ti ; up to 700°C
Ni–alloy: Fuel sheath; up to 700–750°С EP–753 00Сr18Ni40Mo
5NbVB <0.01 17.5–19 39–41 4.5–5.0 0.25–0.6 <0.015 0.05–0.2 <0.05 0.8–1.2 EP–684 0Сr20Ni47Mo
8NbCu
0.08 20–22 47–49 8–9 0.7–1.2 <0.005 <0.1
dpa – displacement per atom
2. SUPERCRITICAL-WATER EXPERIMENTAL SETUP FOR OUT-OF-PILE OPERATION
Research Centre Rez Ltd. built the SCWL as a part of an European joint research project, called High Performance Light Water Reactor Phase 2 (HPLWR Phase 2) co-funded by the European Commission [7]. The loop is designed for operation inside a research reactor LVR-15 (Light Water Reactor) in the Research Centre Rez Ltd. The major objective is using this experimental facility for corrosion studies of materials, studies of radiolysis in supercritical water and water chemistry optimization. The research reactor LVR-15 is located at the Research Centre Rez Ltd. The reactor LVR- 15 is the light-water moderated and cooled tank reactor with forced-circulation cooling system.
The following subsections summarize a basic design of the testing loop, especially, focusing on the irradiation channel, which will be inserted inside the LVR-15 reactor core in the future for an in-pile testing of the loop. The SCW loop will widen today's knowledge mainly in material specifications before and after irradiation conditions and radiolysis model from sub-critical to supercritical conditions. In the future, the loop will also be equipped with the special loading device in order to irradiate samples inside the channel for the Stress Corrosion Cracking (SCC) tests. The following groups of materials may be considered for testing: (i) austenitic stainless steels, (ii) ferritic/martensitic (F/M) steels and (iii) Ni-alloys;
several steels were ODS (Oxide Dispersion Strengthened). However, ODS steels are under development, and Ni-based alloys can be unacceptable for core structures due to high parasitic absorption of nickel. Water-chemistry regimes chosen for tests are the Normal Water Chemistry (NWC) simulated in autoclaves by additions of ~150 ppb oxygen, and Hydrogen Water Chemistry (HWC) – pure water with ~30 cc/kg hydrogen [8].
2.1. SCW Loop Overview
The SCWL is a closed loop with forced circulation. It consists of the irradiation channel (which is to be located inside the LVR-15 reactor core) and auxiliary circuits next to the reactor in the reactor hall. The temperature and pressure in the testing area of the irradiation channel are 600°C and 25 MPa, correspondingly. Due to construction materials limitations for nuclear experimental devices, where the maximum surface temperature of material is 500°C for austenitic stainless steel, the target parameters inside the SCW irradiation channel can be achieved only in a restricted volume of the test section. The SCW loop consists of the following auxiliary circuits: Primary circuit (Irradiation channel, Regeneration exchanger, Cooler, Main circulating pump, Electrical heater, and Pressurizer), Purification system, On- line water-chemistry measurement system, Sampling system, Dousing system, Cooling circuits (secondary and tertiary), System of organized leakage, and, Control system and electrical installation.
2.2. Irradiation Channel Description
The main part of the newly designed and built SCWL in the Research Centre Rez Ltd. is its irradiation channel. This channel is composed of a pressure tube build in aluminium
“spacer”. The space between the pressure tube and this spacer is filled with air and it serves as isolation between the SCWL pressure tube at ~450°C and water inside the reactor primary circuit at ~50°C. The irradiation channel has variable internal dimensions according to the actual experimental program, which is a great advantage of this design.
The channel internals are formed by several coaxial tubes (providing the flow), regeneration exchanger, electrical heater, isolation elements and numerous fixing and
stabilizing elements. The regeneration exchanger is placed above the sample area and is used to reach a high temperature inside that area. It is made of 36 stainless steel tubes 3 × 0.2 mm;
six of these tubes are occupied by wiring of the electrical heating (see Fig.5). This exchanger serves as both cooling and heating element in order to achieve a reasonable temperature at inlet and outlet of the irradiation channel. The inlet water serves as a coolant for the external pressure tube to maintain the temperature below 450°C. Then, the regeneration exchanger together with complementary electrical heating and eventually γ-heating while the reactor is in operation, serve to achieve the desired temperature inside the sample area, i.e., 600°C.
After leaving the sample area, the regeneration exchanger re-cools water to an acceptable temperature before it exits the channel. Additional heat source inside the irradiation channel is a specially designed electrical heater, which power is limited by the available space inside the channel to ~15 kW.
Figure 5. Cross-section of the irradiation channel in the location of the regeneration exchanger and regeneration exchanger design [7].
2.3. Performed Tests Phase I- trial period
In the first phase of testing the SCW loop the main goal was to achieve operating parameters inside the loop and to compare calculated and simulated parameters with actual measurements. In the test section inside the irradiation channel, where the maximum parameters should have been reached, temperature of 550°C at 25 MPa was achieved [8].
The maximum design temperature of 600°C was not achieved due to the electrical heating, which was set to operate at 50% of nominal power, because of its malfunctioning limitations.
Nevertheless, it has been proven that the expected maximum temperature of 600°C can be achieved with the available electrical power.
Phase II- loop non-active operation
The second phase of SCW testing was divided into two parts: 1) primary circuit testing and 2) irradiation channel testing.
The first tests were performed without internals of irradiation channel (dummy channel), where thermohydraulic losses were simulated. Operating parameters for the primary
circuit at the entrance to the irradiation channel were reached, i.e., the pseudocritical point at 25 MPa was overcame by several degrees. The tests revealed minor flow instability and consequently temperature instability in the primary circuit. Consequently, a control valve was re-designed in order to improve regulating characteristics; and after repeating the tests, no flow instabilities occurred.
For the second test, the designed internals were inserted into the irradiation channel, and a new headpiece with wiring was attached to thermocouples and heating elements were connected to the channel [8]. This test had an objective to achieve the highest temperature in the channel focusing on water flow through the channel’s internal parts and temperature distribution. For reaching required temperature, water-flow optimization had to be performed by splitting the flow. Temperature distribution inside the channel is shown in Fig. 6. The heating power had to be limited due to unresolved problems with heating power control system. However, there was a very good agreement between calculated and measured data.
Other measured parameters during this testing are presented in Fig. 7.
Figure 6. Temperature distribution in irradiation channel – comparison of calculated and measured values* [8].
* Marked points represent real data measured during tests. Circle points represent temperature distribution at 200 kg/h flow rate through the channel and triangular points represent temperature distribution at flow splitting and with heating power reserve.
Figure 7. Supercritical parameters in the SCWL during the non-active testing.
3. CONCLUSIONS
The SCWL in the Research Centre Rez Ltd. is a unique facility, actually, one of the first such facilities in the world. Performed tests and obtained results from this loop will help in the future in understanding the supercritical state of water and utilization of that medium inside a nuclear reactor. Supercritical water is widely used in coal-fired thermal-power plants, but until today no one nuclear reactor operates with it. Therefore, research in this area, especially, building and operating the experimental device, which can be tested inside the reactor, will also help to understand better supercritical-water reactors as part of Generation- IV concepts. Planned tests in the SCW loop within the area of corrosion, radiolysis and water chemistry in supercritical state of water will give enormous amount of data, which are necessary for designing and licensing SCW reactors in the future.
Recently, the SCW loop irradiation channel was designed and tested at Research Centre Rez Ltd. This step was a big challenge considering the lack of information in this field, and technical limitations inside the loop itself such as space, maximum temperature, flow rate, etc., plus also very strict safety issues. After the first out-of-pile tests, we were able to justify a proper design of the irradiation channel, and it was verified that design parameters (P = 25 MPa and T = 600°C) will be reached in the test section during in-pile testing. In the future it is planned that out-of-pile testing will continue for achieving the designed temperature of 600°C, and after experiments with specific specimens of materials will be performed, consequently leading to the in-pile testing of the SCW loop inside the LVR-15 research reactor.
REFERENCES
[1] Pioro, I.L. and Duffey, R.B., Heat Transfer and Hydraulic Resistance at Supercritical Pressures in Power Engineering Applications, ASME Press, New York, NY, USA, 2007, 334 pages.
[2] Naterer, G., Suppiah, S., Lewis, M., & Pioro, I. (2009). Recent Canadian advances in nuclear-based hydrogen production and the thermochemical Cu-Cl cycle. Int. J. of Hydrogen Energy (IJHE), 34, 2901-2917.
[3] Saltanov, Eu. and Pioro, I., 2011. World Experience in Nuclear Steam Reheat, Chapter in book “Nuclear Power - Operation, Safety and Environment”, Editor: P. Tsvetkov, INTECH, Rijeka, Croatia.
[4] Oka, Y, Koshizuka, S., Ishiwatari, Y., and Yamaji, A., 2010. Super Light Water Reactors and Super Fast Reactors, Springer, 416 pages.
[5] Kruglikov, P.A., Smolkin, Yu.V. and Sokolov, K.V., 2009. Development of engineering solutions for thermal scheme of power unit of thermal power plant with supercritical parameters of steam, (In Russian), Proc. Int. Workshop "Supercritical Water and Steam in Nuclear Power Engineering: Problems and Solutions”, Moscow, Russia, October 22–23, 6 pages.
[6] Rodchenkov, B.S. and Yarmolenko O.A., 2009. Constructional materials for graphite- moderated reactors on supercritical parameters, (In Russian). Proc. Int. Workshop
"Supercritical Water and Steam in Nuclear Power Engineering: Problems and Solutions”, Moscow, Russia, October 22–23, 16 pages.
[7] Ruzickova, M., Hajek, P., Smida S., Vsolak, R., Petr, J., Kysela J. “Supercritical water loop design for corrosion and water chemistry tests under irradiation”, The 3rd Int.
Symposium on SCWR – Design and Technology, March 12-15, 2007, Shanghai, China.
[8] Ruzickova, M., Vsolak, R., Hajek, P., Zychova, M., and Fukac, R. The supercritical water loop for in-pile materials testing”, The 5th Int. Sym. SCWR (ISSCWR-5) Vancouver, British Columbia, Canada, March 13-16, 2011.
