Atomic Energy, Vol. 116, No. 5, September, 2014 (Russian Original Vol. 116, No. 5, May, 2014)
ARTICLES DEVELOPMENT OF SUPERCRITICAL-WATER COOLED REACTORS IN RUSSIA AND ABROAD
A. P. Glebov and A. V. Klushin
UDC 621.311.25 + 621.039.51
Analysis of foreign research on SCWR attests to their systemacy and consistency and, especially, the timely training of cadres. The knowledge accumulated in our country in the last 10 years makes it possible to refine the previously developed concept of the 1700 MW(e) VVER-SKD reactor, indicate a plan for top-priority research, formulate a technical mission, and proceed to the design of an experimental 30 MW reactor.
Two international conferences devoted to innovative reactors were held in China in 2013. Nearly 100 reports concerning reactors cooled by supercritical water were presented at the first one held in Shenzhen. The second conference, held in Chengdu in July–August, was organized into 17 sections in which the main questions concerning NPPs were examined – operation, protection, safety, thermohydraulics, materials and fuel cycles. Special attention was devoted to the training of young specialists. It can be concluded from the reports presented at these conferences that nuclear power is developing rapidly, including reactors with supercritical coolant parameters in Asia. Aside from Japan and South Korea, China leads in terms of the rates of construction of new NPP. If nuclear power production in China now amounts to 1.8% (12.9 GW), its contribution will increase to 18% (60 GW) by 2020. Development of Water-Cooled Reactors with Supercritical Coolant Parameters. Research in this area is now being conducted in 15 countries (Table 1). For foreign specialists, the top-priority task is to develop a reactor with a thermal neutron spectrum and then switch to a reactor with a fast neutron spectrum. The Japanese project, whose funding started in 2000, is adopted as the baseline [1]. The SCLWR core is cooled by water under pressure 25 MPa, temperature 280°C at the entrance and 508°C at the exit and consists of 96 300 × 300 mm square fuel assemblies (Fig. 1). A fuel assembly consists of 301 fuel elements and 36 square rods with water, in 16 of which directed tubes of the safety-and-control system are positioned. The feed water, entering from top to bottom, cools the reactor vessel and fills the rods with water. The flows are mixed at the bottom of the core and, rising in the cavities with the fuel elements, are heated. Such a reactor makes it possible to greatly improve the economic performance of NPP, but this requires enriched uranium and a large amount of spent fuel is produced. In addition, it will not meet the requirements for closing the NFC. Japanese scientists have proposed a design for a reactor with a fast neutron spectrum (SCFR) [2] and fuel based on oxides from depleted uranium and plutonium-enriched uranium (Fig. 2). The core is comprised of 270 fuel assemblies with UO2 + PuO2 fuel and 163 fuel assemblies in the blanket zone with depleted-uranium fuel UO2. The reactor vessel and blanket zone are cooled by 280°C water flowing from top to bottom under pressure 25 MPa. These flows mix with one another at the bottom and in the ascending channel cool the core with coolant temperature 523°C at the exit. The closely spaced lattice of fuel elements and fast neutron spectrum, varying considerably along the height in the core, with blanket zones make it possible to obtain a fuel breeding ratio greater than 1 (BR = 1.034). To level out the power-release field, the fuel enrichment in the
State Science Center of the Russian Federation – Leipunskii Institute for Physics and Power Engineering (GNTs RF – FEI), Obninsk. Translated from Atomnaya Énergiya, Vol. 116, No. 5, pp. 267–274, May, 2014. Original article submitted December 16, 2013.
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1063-4258/14/11605-0320 ©2014 Springer Science+Business Media New York
TABLE 1. Characteristics of Water-Cooled Reactors with Supercritical Coolant Parameters SCWR (Korea)
SCLWR (Japan)
CANDU+ (Canada)
HPLWR (European Union)
SCWR (China)
SCFR (Japan)
VVER-SKD (Russia)
thermal
3989
2273
2540
2188
2284
3832
3830
electric
1739
950
1220
1000
1000
1698
1700
Efficiency, %
43.7
42
48
44
43
44.3
43.5
water
350
280
350
280
280
280
290
steam
510
508
625
508
500
523
540
25
25
25
25
25
25
25
2518
1816
1312
1113
1177
1897
1880
Parameter
Power, MW:
Temperature, °C:
Steam pressure, MPa Water flow rate, kg/sec
Fig. 1. Cartogram of SCLWR fuel assembly with thermal neutron spectrum: 1) water rod; 2) safety-and-control rod; 3) fuel element; 4) channel.
top half of the core must be increased, which results in a positive effect of reactivity when the core is dewatered. To obtain a negative void coefficient of reactivity, it is proposed that blanket fuel assemblies with depleted uranium be introduced and a moderator ZrH1.7 be introduced into the outer layer of the blanket fuel assembly in order to increase neutron absorption. These two reactor concepts were subsequently elaborated and improved, which was reflected in the reports at the conference. For example, for the China variant of a fuel assembly in a reactor with a thermal neutron spectrum it is proposed that two or three rows of fuel elements be arranged around the ‘water columns’ in order to increase the specific power [3]. The European variant of the reactor uses a three-course cooling scheme with the coolant motion starting in the central zone with ascending flow followed by a turn, descending motion in the middle zone and ascending motion in the peripheral zone. Thus the heating of the coolant is divided into three sections [4]. Canada is developing a vertical 300 MW(e) CANDU+ reactor with 120 process channels with coolant temperature 350°C at the entrance and 625°C at the exit with pressure 25 MPa (Fig. 3). A 1220 MW(e) reactor with efficiency 50% will be developed at the second stage [5]. For a reactor with a fast neutron spectrum, the number of fuel assemblies, the power and the arrangement of the fuel assemblies of the blanket in the form of separate fuel assemblies or seven fuel assemblies bundled together, rather than con321
Fig. 2. Arrangement of SCFR fuel assemblies (Japan) with a fast neutron spectrum: 1) fission zone; 2) blanket.
Fig. 3. Vertical CANDU+ reactor (Canada).
tinuous concentric zones, are varied with the overall scheme remaining unchanged [6]. With this arrangement of the fuel assemblies, the blanket will not deform the power release as much. The water-cooled reactor with supercritical coolant parameters which is under development in our country relies on VVER operating experience, many years of global experience in operating thermal power plants, where water vapor with supercritical parameters (25–37 MPa, 540–700°C and 620–700°C) is used, as well as the development of nuclear power facilities with nuclear superheated steam and experience in operating steam-superheating fuel elements at the Beloyarskaya NPP (540–560°C). In Russia, the Leipunskii Institute for Physics and Power Engineering (FEI), Gidropress Experimental Design Bureau (OKB Gidropress), and the National Research Center Kurchatov Institute jointly developed the VVER-SKD reactor concept with two variants of the core with thermal and fast neutron spectra and different schemes for removing heat. The main characteristics, construction, materials and configurations of the power block were determined and questions concerning safety, water chemistry and others were examined. Work is currently proceeding on a single-loop nuclear power facility with supercritical coolant parameters, a two-course cooling scheme, fast-resonance neutron spectrum and power 1700 MW(e) [7]. In accordance with the proposed cooling scheme, the core is divided along the radius into central and peripheral zones with approximately the same number of fuel assemblies 121 and 120, respectively (Fig. 4). The peripheral zone is 322
Fig. 4. Cooling scheme for the VVER-SKD reactor: 1) safety-and-control rods; 2) cap; 3) inner cap; 4) body; 5) heat insulation; 6, 7) out and in connecting piece, respectively; 8) core; 9) shaft; 10, 11) fuel assemblies of the descending and ascending sections, respectively; 12) isolation shell.
cooled with the coolant flowing from top to bottom. The coolant flows from the peripheral fuel assemblies are combined at the bottom of the core in a mixing chamber and enter the central zone, which is cooled with the coolant flowing from bottom to top. In the descending section, the coolant will be heated to 385°C and the density will change by a factor of approximately 3. In the ascending section, the coolant will be heated to 540°C, and the density will change by a factor of 2.2. Thus, the neutron spectrum changes little over the height, and because of the change along the radial direction there is no need for complicated shaping of the fuel enrichment for purposes of equalizing the energy release over the volume of the core; the void effect will be negative without the introduction of a blanket. All fuel-assembly designs, as compared with a single-course cooling scheme, will operate with half the temperature differential. In connection with a 10-fold reduction of the coolant flow rate compared with VVER, the coolant velocity will be low and the losses to friction will be ~1.2 MPa. As the coolant velocity increases by a factor of 2, the heat-emission coefficient will increase by a factor of 1.7, as a result of which the temperature of the fuel-element cladding will decrease and the serviceability of the cladding will increase. This reactor is regarded as a promising development of VVER with the possibility of transitioning to a closed fuel cycle [8, 9]. The VVER-SKD reactor has the following advantages: 1) high-energy (fast-resonance) neutron spectrum, making it possible to attain the working fuel breeding ratio (~1), reduce the consumption of uranium, and permit the use of 238U with Am, Np, and Cm being burned out; 2) efficiency of a cycle increased to 44–45% instead of 33–34%; 3) reduced coolant flow through the core, permitted by the possibility of increasing the coolant heating in the core by 250°C as compared with 30–35°C; 4) straight-through scheme of NPP, making it possible to do without second-loop equipment, such as steam generators, pumps and other; 5) use of well-understood serially produced equipment in the machine room (turbines, heaters); 6) simplification of safety systems compared with VVER, enhanced hydrogen safety with elimination of zirconium alloys; 7) significant reduction of the volume of the containment shell and amount of construction work; 323
Fig. 5. Cartogram of the core of a VVER-SKD reactor: 1, 2) fuel assemblies in the central and peripheral zones, respectively.
8) lower specific metal-intensiveness; this index equals 17.7 for BN-350, 13 for BN-600, 9.7 for BN-800, 3.25 VVER-1000, and 1.5 tons/MW for VVER-SKD, respectively; and 9) lower operating expenses. Scientific and Technical Solutions for the Development of VVER-SKD. To validate and design the experimental 30 MW(t) reactor, it is necessary to determine the acceptable coefficients of nonuniformity of power release in the core, the fuel breeding ratio BR and the burnup and to secure negative coefficients of reactivity in different operating regimes. The vessel and fuel-element cladding materials must be picked on the basis of the experience gained in operating BOR-60, BN-350, and BN-600. These are likely to be chromium-nickel alloys (17Cr13Ni2Mo), and the irradiation intensity in VVER-SKD will be 2–3 times lower than in a fast reactor. The particulars of thermohydraulic processes in such a reactor are a significant change in water density and specific heat with supercritical parameters (critical point temperature 374°C and pressure 22 MPa). The dependences of the heat-emission coefficient and the hydraulic resistance have been studied above and below the critical pressure for channels with different forms. These characteristics are more difficult to measure for near-critical parameters. At present, the uncertainty (error) in the calculations of the heat-emission coefficients for channels with simple forms (round tube, planar gap) is about ±15%. This uncertainty (error) was adequate at the initial phase of the development work, but additional experiments performed in channels with simple and complicated forms, first and foremost, for bundles of rods with close packing, are needed. The mixing processes for streams of ‘cold’ (temperature below critical) and ‘hot’ (temperature above critical) water must be studied in order to determine the unstable regimes. The specifics of using water with supercritical parameters in nuclear reactors are related with the action of radiation and the accompanying radiolysis. This phenomenon must be evaluated and experimentally investigated in a high-energy neutron spectrum, including the mass-transfer of the products of corrosion and radionuclides. The safety systems in VVER-SKD are similar to those used in VVER-1000. A particularity of the experimental VVER-SKD-30 and VVER-SKD could be the organization of natural circulation in the loop by means of special setups or individual channels for removing heat. Software, including codes for making better evaluations, for performing coupled calculations of the neutron-physical and thermohydraulic characteristics of VVER-SKD, which will take account of the complexity of the change in the properties of the coolant in the working temperature range at the entrance into and exit from the reactor 350 and 550°C, respectively, must be developed and verified. VVER-SKD-30 Experimental Reactor. The particulars of the 30 MW(t) VVER-SKD-30 experimental reactor cooled by water with supercritical parameters (pressure 25 MPa, temperature at the entry and exit 290 and 540°C, respectively) were examined in a report by specialists from the FEI [10]. The reactor is characterized by a fast-resonance neutron 324
spectrum and a two-course scheme for cooling the coolant. The use of different fuel loads and schemes for reloading fuel assemblies in the course of fuel burnup was examined. The characteristics of the variants of the core of the VVER-SKD-30 experimental reactor are as follows: Fuel UO2 Core, cm: deq 73.9 h 85 Number of fuel assemblies in the core: central zone 91 peripheral zone 90 Fuel-assembly spacing, cm 5.23 Number of fuel elements per fuel assembly: central zone 19 peripheral 18 Fuel-element spacing, mm 12 Fuel-element cladding, mm: diameter 10.7 thickness 0.55 Mass loaded into the reactor, kg: fuel 1728.4 fissile isotopes 345.7 Average enrichment with fissile isotopes, % 20 Fuel-assembly run time, eff. days 4 × 270 Power generation by off-loaded fuel assemblies, MW·days/kg h.a.: average 19 maximum 27.2 Reactivity excess, % 0.93 Maximum coefficient of nonuniformity of power release: Kq 1.18 Kv 1.59 Breeding ratio (average over core) 0.89 Maximum neutron flux density, 1014 sec–1·cm–2: fast (E ≥ 0.11 MeV) 3.91 total (E ≥ 4 eV) 6.63 82.4 Average specific energy density of the core, W/cm3 Average heat flux from the surface of fuel elements, W/cm 105.4
(U+Puweap)O2
(U+Pupower)O2
66.1 70
66.1 70
73 72 5.23
73 72 5.23
19 18 12
19 18 12
10.7 0.55
10.7 0.55
1139.2 256.6 20.36 4 × 250
1159.7 236.1 21.9 4 × 260
29.6 36.7 1.38
32.3 43.3 1.73
1.22 1.60 0.9
1.22 1.63 0.9
5.83 9.87 125 159.7
5.18 8.85 125 159.7
The objectives of the project are to develop VVER-SKD concepts for design solutions for the core and reactor, test and validate the material picked and the water-chemistry and nominal regimes and investigate safety. VVER-SKD for Use in Closed Fuel Cycles (Figs. 5 and 6). The VVER-SKD fuel is a mixture of spent VVER nuclear and weapons grade plutonium. For effective density of the mixture of uranium and plutonium oxides 9.5 g/cm3, the density of the weapons-grade plutonium oxide is 0.7 g/cm3. The calculations were performed using a domestic software system WIMS-ACADEM in a five-group approximation for a three-dimensional hexagonal geometry. In addition to uranium-plutonium fuel, the possibility of bringing thorium into the mixed loads was also examined: uranium-plutonium in the central zone, uranium-thorium in the peripheral or uranium-thorium in the entire core (Table 2) [11]. It is evident from the computational results presented that because of the characteristics of the reactor, viz., fast-resonance neutron spectrum and two-course cooling scheme with denser coolant in the peripheral zone, the void coefficient of 325
Fig. 6. Transverse section of fuel assembly: 1) jacket with thickness 2.25 mm; 2) 12 × 0.55 mm central tube; 3) 18 guiding channels under an absorbing element of size 12 × 0.55 mm; 4) 252 fuel elements, 10.7 × 0.55 mm cladding, spacing 12 mm.
reactivity is negative throughout the entire run. When the reactor is filled with cold water, it will be necessary to use absorbing rods with enriched boron, but even in this case gadolinium must be introduced in the variant with a uranium-thorium fuel load. The first calculations assumed jacketed fuel assemblies as well as an isolation barrier to prevent leakage between fuel assemblies in the peripheral and central zones (see Figs. 4–6). The bulk of the calculations neglected the isolation barrier. Different design variants of barrier were considered: assembled from hollow steel blocks of size corresponding to the fuel assemblies, or a continuous construction consisting of layers of steel (2 cm), heat insulation (4 cm), and zirconium alloy (4 cm). The isolation barrier increases the enrichment of the fuel and creates a large nonuniformity of the energy release along the fuel assemblies. For example, for the variant of the barrier in the form of a layered structure with the same fuel run 5 × 300 eff. days and number of fuel assemblies the enrichment with weapons grade plutonium must be increased by 4% (to 0.73 eff. days). But, if the barrier is present, it is possible to switch to jacketless fuel assemblies, as is done in VVER-1000, which improves heat exchange and decreases the maximum temperature of the coolant at the exit. A fuel load differing from the initial load (neglecting the barrier) was examined: jacketless cassettes with spacing 205 instead of 207 mm, number of fuel elements, unchanged lattice spacing, 0.6 g/cm3 plutonium in the central zone in 25 central fuel assemblies and all other fuel assemblies (central and peripheral zones) 0.7 g/cm3 with no changes, the number of fuel assemblies in the peripheral zone is taken to be 114 instead 120 and in the central zone with no changes 121, the isolation barrier was simulated by adding in the peripheral zone 18 fuel assemblies with material corresponding to a barrier with a homogeneous composition. The calculations showed that jacketless fuel assemblies make it possible to reduce the consumption of weapons grade plutonium by 5% and at the same time increase the fuel-assembly run by 250 eff. days. The breeding ratio does not change and on average over the core equals 0.94. The neutron flux density at the center of the core will increase by a factor of 1.5 and comprise at the end of the run 1.14·1015 sec–1·cm–2 (E ≥ 0.11 MeV) with the total being 2.26 cm 1015 sec–1·cm–2. The VVER-SKD reactor can be used effectively in a closed fuel cycle because it operates on its own spent nuclear fuel with a small amount of added plutonium, weapons-grade or without the blankets of fast reactors. Uranium-plutoniumthorium fuel can also be used in it. The main question in managing spent fuel becomes managing 241–243Am and 242–245Cm. The isotope 237Np is not separated from the fuel and is shared with it. Research on heterogeneous incineration of americium and curium in the fuel of 326
TABLE 2. Main Characteristics of a Reactor with Uranium-Plutonium-Thorium Fuel Cycle Parameter
U–Pu
Pu–Th
Th
135.6
137.3
139
11.77/0
5.91/4.8
0/10.8
48.86/0
48.86/39.99
50.24/39.46
central zone
7.7/0
7.7/0
0/9
peripheral zone
7.7/0
0/7
0/6.9
5
5
5
300
310
300
average
39.79
42.2
34.6
maximum
65.4
68.6
47.5
Kq
1.46
1.61
1.67
Kv
2.19
2.62
2.8
Load of fissile isotopes, tons/yr
2.34
2.11
2.2
Unloading of fissile isotopes, tons/yr
2.18
1.87
1.96
central zone
1.013
1.003
0.957
peripheral zone
0.853
0.769
0.8
average over the core
0.933
0.887
0.89
run start
–5.88
–3.24
–6.28
run finish
–3.64
–1.4
–2.32
Initial fuel load, tons 233
Initial load of fissile isotopes Pu/
U in the core, tons
Load of fissile Pu/233U in fuel assembly, kg 233
Fuel enrichment Pu/
U, %:
Refueling number Time between refueling, eff. days Energy production, MW·days/kg h.a.:
Maximum coefficient of nonuniformity of energy release:
Breeding ratio:
Dewatering ΔK, %:
some fuel elements in working fuel assemblies in application to BN-1200 has shown that for a nominal run of five calendar years only 45% will be burned up, and therefore such fuel elements should be included in the subsequent recycle [12]. Deeper burnup of actinides can be attained in VVER-SKD, if the fuel assemblies containing it are placed in the peripheral zone for two 5-yr runs [13]. In this time, 1400 kg of transuranium elements, of which Am amounts to 97% and Cm to 3% (Fig. 7), will accumulate in the stationary regime. Calculations of the fuel cycle with a stationary refueling regime and a 5-yr run time were performed for working and 10-yr run time for fuel assemblies with actinides. The fuel assemblies Nos. 15, 28, and 39 simulate in the calculations the presence of an isolation barrier and correspond to its volume and composition (Fig. 8). In fuel elements with actinides, 12% of the initial content remains after 10 yr burnup, and such fuel elements can be placed in long-term storage. Twenty-four fuel assemblies are loaded with 1230 kg 241–243Am. Thus, the actinides which have accumulated in the reactor over 10 years of operation can burn up, which requires 24 fuel assemblies. This makes it possible to conserve 40 kg of plutonium per year without any great changes in the nonuniformity of the distribution of the energy release over the core. 327
Fig. 7. Transverse section of a fuel assembly with Am and Cm: 1) central tube; 2) tubes beneath the absorbing rods of the safety-and-control system; 3, 4) fuel elements with mixed fuel and americium and curium, respectively.
Fig. 8. Energy production and nonuniformity of energy release over the fuel assemblies in a core with Np, Am, and Cm in a corner with symmetry 60° at the end of a run: 1) fuel assembly No.; 2) burnup, Mw·days/kg h.a.; 3) nonuniformity of energy release; a) fuel assembly with actinides.
Conclusion. Analysis of foreign research shows that it is systematic and consistent and, most importantly, that the training of personnel is timely. Foreign institutes conduct together with the IAEA special courses on the construction and technology of SCWR. International symposia and meetings of working groups, where practical questions are discussed, the latest advances in reactor physics, materials, water chemistry and corrosion, thermohydraulics, safety and thermal schemes of NPP and other questions are covered, are held for specialists who are already working in this field of nuclear engineering. The knowledge and experimental resources accumulated in the last 10 years in our country make it possible to refine the previously developed concept, indicate plans for top-priority research, formulate a technical mission and begin design work on a small, experimental 30 MW(t) reactor. 328
REFERENCES 1. 2. 3. 4. 5. 6. 7. 8. 9. 10.
11.
12. 13.
Y. Oka and S. Koshizuka, “Design concept of once-through cycle supercritical-pressure light water cooled reactors,” Proc.1st Int. Symp. Supercritical Water-Cooled Reactors, Japan, Nov. 6–9, 2000, Rep. No. 101, pp. 1–22. J. Wu and Y. Oka, “Core design of super LWR with double tube water rods,” ISSWCR-6, China (2013), Paper 13009. W. Zang, “Preliminary core conceptual design of 1000 MWe SCWR,” ibid., Paper 13066. T. Schulenberg, “Supercritical water-cooled reactor (SCWR),” Report on the Working Group IAEA on SCWR Programme, Canada, Sept. 19–23, 2011. M. Yetisir. M. Gaudet, and D. Rhodes, “Development and integration of Canadian SCWR concept with counter-flow fuel assembly,” ISSWCR-6, China (2013), Paper 13059. B. Lui, L. Cao, H. Wu, and Y. Zheng, “Three-dimensional core analysis on the breeding capability of the super fast reactor,” ibid., Paper 13121. A. P. Glebov and A. V. Klushin, “Reactor with a fast-resonance neutron spectrum, cooled by supercritical pressure water with a two-course coolant motion scheme,” At. Énerg., 100, No. 5, 349–355 (2006). Yu. D. Baranaev, A. P. Glebov, A. V. Klushin, and V. F. Ukraintsev, “Use of a reactor cooled by supercritical pressure water – VVER-SKD in a closed fuel cycle,” Izv. Vyssh. Uchebn. Zaved. Yad. Energet., No. 3, 18–31 (2010). Yu. D. Baranaev, A. P. Glebov, P. L. Kirillov, and A. V. Klushin, “Reactor cooled by supercritical pressure water VVER-SKD – the main candidate for Super-VVER,” Preprint FEI-3188 (2010). Y. D. Baranaev, A. P. Glebov, P. L. Kirillov, and A. V. Klushin, “Neutronic characteristics of a 30 MWt SCW experimental reactor: from water-cooled power reactor technology to a direct cycle nuclear reactor with supercritical water parameters and fast neutron spectrum,” ISSWCR-6, China (2013), Paper 13108. A. P. Glebov, A. V. Klushin, Y. D. Baranaev, and P. L. Kirillov, “Presearch of features of U–Pu–Th fuel cycle and its use for burning up of minor actinides in supercritical watercooled reactor with fast neutron spectrum,” ICONE-21, China (2013), Paper 16888. V. M. Poplavskii, A. M. Tsibulya, Yu. S. Khomyakov, et al., “Core and fuel cycle for a promising sodium reactor,” At. Énerg., 108, No. 4, 206–211 (2010). Yu. D. Baranaev, A. P. Glebov, and A. V. Klushin, Patent 2485612 RF, “Core with fast-resonance neutron spectrum with supercritical pressure water,” March 5, 2012, Byull. Izobret. Polezn Modeli, No. 17, 45 (2013).
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