Abstract
The article presents information on the validation and verification (V&V) of the first version (V1) of the EUCLID integrated code intended for safety analysis of operating or designed liquid metal (sodium, lead, or lead–bismuth) cooled reactors under normal operation and under anticipated operational occurrences by carrying out interconnected neutronics, thermal–mechanical, and thermal–hydraulic calculations. The list of processes and phenomena that have to be modeled in the integral code for correctly describing the above-mentioned operating conditions is given. Based on this list, the most high-quality experimental data are selected for carrying out the validation. It is shown that, for sodium cooled reactors, a significant number of experiments was carried out around the world on studying individual thermal–hydraulic processes and phenomena, which made it possible to perform validation of the thermal–hydraulic module. The validation of the code—as applied to description of processes that take place in fuel rods with oxide or nitride fuel and gas gap—is carried out against the results of post-pile investigations of fuel rods irradiated in fast sodium cooled research and power-generating reactors. The obtained results opened up the possibility to determine the errors of calculating such fuel rod parameters as release of gaseous fission products from the fuel and sizes of pellet and cladding in a limited range of burnup values. To perform validation of the neutronics module as applied to calculation of such parameters as power density distribution over the core and decay heat release, a sufficient number of experiments and benchmarks were selected. The results obtained from experimental operating conditions of a BN-600 reactor and startup conditions of a BN-800 reactor made it possible to estimate how correctly the integral code performs calculations of interconnected thermal–hydraulic and neutronic processes. Only a limited set of experimental investigations is available for heavy liquid metal cooled reactors. In view of this circumstance, programs for obtaining the lacking data are developed. To estimate the quality with which the experiments are modeled by means of the EUCLID/V1 integrated code, a procedure for evaluating the errors of calculation results is developed. In accordance with this procedure, the error of calculating the parameters playing the main role in the reactor safety assessment is evaluated.
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References
White Book of Nuclear Power Generation, Ed. by E. O. Adamov (NIKIET, Moscow, 2001) [in Russian].
P. N. Alekseev, V. G. Asmolov, A. Yu. Gagarinskii, N. E. Kukharkin, Yu. M. Semchenkov, V. A. Sidorenko, S. A. Subbotin, V. F. Tsibul’skii, and Ya. I. Shtrombakh, “On a nuclear power strategy of Russia to 2050,” At. Energy 111, 239–251 (2011).
L. A. Bol’shov, N. A. Mosunova, V. F. Strizhov, and O. V. Shmidt, “Dedicated to the 60th anniversary of the journal Atomnaya Energiya: Next generation design codes for a new technological platform for nuclear power,” At. Energy 120, 369–379 (2016).
A. V. Avvakumov, V. M. Alipchenkov, A. A. Belov, V. P. Bereznev, A. V. Boldyrev, N. A. Grushin, I. N. Khanbikov, I. A. Klimonov, P. V. Kolobaeva, D. A. Koltashev, N. A. Mosunova, V. D. Ozrin, N.A.Rtishchev, E. F. Seleznev, M. M. Semenova, A. A. Stakhanova, V. F. Strizhov, V. I. Tarasov, E. V. Usov, D. P. Veprev, V. A. Veretentsev, D. A. Afremov, A. V. Kudryavtsev, A. A. Semchenkov, S. L. Osipov, A. M. Anfimov, and V. S. Gorbunov, “Coupled calculations for the fast reactors safety justification with the EUCLID/V1 integrated computer code,” in Proc. Int. Conf. on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17), Yekaterinburg, Russia, June 26–29, 2017, paper no. IAEA-CN245-184.
N. A. Mosunova, “The EUCLID/V1 integrated code for safety assessment of liquid metal cooled fast reactors. Part 1: Basic models,” Therm. Eng. 65, 69–84 (2018). doi 10.1134/S0040601518050063
RD-03-34-2000. Requirements to the Contents of a Report on Verification and Substantiation of Software Solutions Used for Safety Substantiation of Nuclear Power Objects, Put in Action by Order No. 122 of Gosatomnadzor Rossii of December 28, 2000.
O. D. Kazachkovskii, G. K. Antipin, V. A. Afanas’ev, V. F. Bai, V. A. Borisyuk, E. V. Borisyuk, V. M. Gryazev, V. N. Efimov, V. P. Kevrolev, V. I. Kondrat’ev, N. V. Krasnoyarov, and A. M. Smirnov, “Emergency cool-down of BOR-60 unit,” At. Energ. 34, 341–344 (1973).
I. A. Klimonov, E. V. Usov, G. A. Dugarov, A. A. Butov, I. G. Kudashov, E. N. Ivanov, N. A. Mosunova, V. F. Strizhov, A. M. Anfimov, V. S. Gorbunov, D. V. Kuznetsov, S. L. Osipov, and A. I. Bel’tyukov, “HYDRA-IBRAE/LM/V1 thermohydraulic code verification based on BN-600 experiments,” At. Energy 122, 258–262 (2017).
V. N. Fromzel’, L. V. Fromzel’, and N. V. Vdovets, “The method for determining effective heat conductivity of fuel rod assembly and calculation of temperature field in assemblies placed in vertical containers,” in Processes of Heat and Mass Transfer and Hydrodynamics in Safety Systems of NPPs with VVER-640: Collection of Papers (Tsentr. Kotlo-Turbinnyi Inst., St. Petersburg, 1997), pp. 139–150 [in Russian].
Yu. A. Zeigarnik and V. D. Litvinov, “The study of hydraulic resistance during sodium boiling in a pipe,” Teplofiz. Vys. Temp., No. 5, 1116–1118 (1977).
Yu. A. Zeigarnik and V. D. Litvinov, “Experimental study of heat transfer and pressure losses during sodium boiling in a vertical pipe,” in Proc.5th Conf. on Heat and Mass Transfer, Minsk, 1975, Vol. 3, Part 1, pp. 147–156.
H. Kottowski and C. Savatteri, “Fundamentals of liquid metal boiling thermohydraulics,” Nucl. Eng. Des. 82, 281–304 (1984).
D. N. Wall and A. A. Cooper, “An analysis of the pressure drop and dryout results from the second ISPRA 12-pin gridded cluster,” in Proc. 12th Liquid Metal Boiling Working Group (LMBWG), Ispra, Italy, 15–17 October 1986 (1987), pp. 191–220.
C. Savatteri, R. Warnsing, and H. Kottowski, “Twophase flow pressure drop of boiling sodium in grid and wire-spaced bundles,” in Proc. 13th Liquid Metal Boiling Working Group (LMBWG), Winfrith, UK, Sept. 27–29, 1988 (1989), pp. 99–120.
H. M. Kottowski, C. Savatteri, and W. Hufschmidt, “A new critical heat flux correlation for boiling liquid metals,” Nucl. Sci. Eng. 108, 396–413 (1991).
Y. Kikuchi and K. Haga, “Sodium boiling experiments in a 19-pin bundle under loss-of-flow conditions,” Nucl. Eng. Des. 66, 357–366 (1981).
A. Kaiser and W. Peppler, Sodium Boiling Experiments in an Annular Test Section under Flow Rundown Conditions, Report KFK-2389 (1977).
J. Aberle, A. J. Brook, and W. Peppler, Sodium Boiling Experiments in a 7-Pin Bundle under Flow Rundown Conditions, Report KFK-2378 (1976).
M. H. Fontana, R. E. MacPherson, P. A. Gnadt, L. F. Parsly, and J. L. Wantland, Temperature Distribution in a 19-Rod Simulated LMFBR Fuel Assembly in a Hexagonal Duct Fuel Failure Mockup Bundle 2a) — Record of Experimental Data, ORNL-TM-4113 (Oak Ridge National Lab., Oak Ridge, TN, 1973).
N. Hanus, W. R. Nelson, N. E. Clapp, M. H. Fontana, P. A. Gnadt, R. H. Thornton, and J. L. Wantland, Steady-State Sodium Tests in a 19-Pin Internally Guard-Heated Simulated LMFBR Fuel Assembly with a Six-Channel Internal Blockage-Record of Experimental Data for THORS Bundle 3C, ORNL-TM-6498 Oak Ridge National Lab., Oak Ridge, TN, 1979).
J. L. Wantland, N. E. Clapp, M. H. Fontana, P. A. Gnadt, and N. Hanus, “Dynamic boiling tests in a 19-pin simulated LMFBR fuel assembly,” in Proc. ANS Winter Meeting, San Francisco, CA, 1977 (American Nuclear Society, La Grange Park, IL,1977); Report No. CONF-771109–77.
C. W. Choi and K. S. Ha, “Validation of the finned sodium–air heat exchanger model in MARS-LMR,” Ann. Nucl. Eng. 94, 213–222 (2016).
BN-600 Hybrid Core Benchmark Analysis, IAEA Technical Document No. IAEA-TECDOC-1623 (2010). ISBN 978-92-0-109409-4. ISSN 1011–4289.
BN-600 MOX Core Benchmark Analysis, IAEA Technical Document No. IAEA-TECDOC-1700 (2013). ISBN 978-92-0-139210-7. ISSN 1011–4289.
M. N. Zizin, L. K. Shishkov, and L. N. Yaroslavtseva, Test Neutron-Physical Calculation of Nuclear Reactors Atomizdat, Moscow, 1980) [in Russian].
Benchmark Analysies on the Control Rod Withdrawal Tests Performed during the PHÉNIX End-of-Life Experiments, IAEA Technical Document No. IAEA-TECDOC-1742 (2014). ISBN 978-92-0-105314-5. ISSN 1011–4289.
Evaluation of Benchmark Calculations on a Fast Power Reactor Core with Near Zero Sodium Effect, IAEA Technical Document No. IAEA-TECDOC-731 (1994).
“Japan’s experimental fast reactor JOYO MK-I core: Sodium-cooled uranium-plutonium mixed oxide fueled fast core surrounded by UO2 blanket,” in Int. Handbook of Evaluated Reactor Physics. Benchmark Experiments, March 2009 ed. (Organization for Economic Cooperation and Development. Nuclear Energy Agency, 2009).
Y. Komano, T. Takeda, and T. Sekiya, Improved Few-Group Coarse-Mesh Method for Calculating Three-Dimensional Power Distributions in Fast Breeder Reactor Report No. NEA/NEACRP/L/204 (Nuclear Energy Agency, 1978).
A. Hebert, “A Raviart–Thomas–Schneider solution of the diffusion equation in hexagonal geometry,” Ann. Nucl. Eng. 35, 363–376 (2008).
“JNDC nuclear data library of fission products. Second version,” JAERY 1320 (1990).
O. W. Hermann and R. M. Westfall, “ORIGEN-S: SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and Association source terms,” NUREG/CR-0200, Revision 4, Vol. 2, Section F7 (1995).
V. M. Kolobashkin, P. M. Rubtsov, P. A. Ruzhanskii, and V. V. Sidorenko, Radiation Characteristics of Irradiated Nuclear Fuel Energoatomizdat, Moscow, 1983) [in Russian].
J. L. Yarnell and P. J. Bendt, “Calorimetric fission product decay heat measurements for 239Pu, 233U, and 235U,” NUREG/CR-0349, LA-7452-MS, Informal report, R-3 (1978).
J. L. Yarnell and P. J. Bendt, “Decay heat from products of 235U thermal fission by fast-response boil-off calorimetry,” LA-NUREG-6713, NRC-3 (1977).
J. K. Dickens, T. A. Love, J. W. McConnell, and R. W. Peelle, “Fission-product energy release for times following thermal-neutron fission of 235U between 2 and 14000 s,” Nucl. Sci. Eng. 74,106–129 (1980).
J. K. Dickens, T. A. Love, J. W. McConnell, and R. W. Peelle, “Fission-product energy release for times following thermal-neutron fission of 239Pu and 241Pu between 2 and 14000 s,” Nucl. Sci. Eng. 78, 126–146 (1981).
K. Baumung, “Measurements of 235U fission-product decay heat between 15 s and 4000 s,” KFK-3262 (Kernforschungszentrum Karlsruhe, 1981).
Yu. S. Khomyakov, Candidate’s Dissertation in Physics and Mathematics I.I. Leypunsky Inst. of Physics and Power Engineering, Obninsk, 1994).
“OECD/NEA Benchmarking of thermal-hydraulic loop models for lead-alloycooled advanced nuclear energy systems. Phase 1: Isothermal forced convection case,” NEA/NSC/WPFC/DOC(2012)17 (2012).
W. Ma, A. Karbojian, B. R. Sehgal, and T.-N. Dinh, “Thermal-hydraulic performance of heavy liquid metal in straight-tube and Utube heat exchangers,” Nucl. Eng. Des. 239,1323–1330 (2009).
W. Ma, A. Karbojian, T. Hollands, and M. K. Koch, “Experimental and numerical study on lead-bismuth heat transfer in a fuel rod simulator,” J. Nucl. Mater. 415, 415–424 (2011).
W. Ma, E. Bubelis, A. Karbojian, B. R. Sehgal, and P. Coddington, “Transient experiments from the thermal-hydraulic ADS lead bismuth loop (TALL) and comparative TRAC/AAA analysis,” Nucl. Eng. Des. 236, 1422–1444 (2006).
W. Ma, E. Bubelis, A. Karbojian, and B. R. Sehgal, “Experimental study on natural circulation and its stability in heavy liquid metal loop,” Nucl. Eng. Des. 237, 1838–1847 (2007).
A. Ciampichetti, D. Pellini, P. Agostiny, G. Benamati, N. Forgione, and F. Oriolo, “Experimental and computational investigation of LBE — Water interaction in LIFUS 5 facility,” Nucl. Eng. Des. 239, 2468–2478 (2009).
P. D. Lobanov, E. V. Usov, A. A. Butov, N. A. Pribaturin, N. A. Mosunova, V. F. Strizhov, V. I. Chukhno, and A. E. Kutlimetov, “Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRAIBRAE/LM thermohydraulic code,” Therm. Eng. 64, 770–776 (2017). doi 10.1134/S004060151710007X
J. Pacio, M. Daubner, F. Fellmoser, K. Litfin, L. Marocco, R. Stieglitz, S. Taufall, and T. Wetzel, “Heavy-liquid metal heat transfer experiment in a 19-rod bundle with grid spacers,” Nucl. Eng. Des. 273, 33–46 (2014).
N. I. Alekseev, S. N. Bol’shagin, E. A. Gomin, S. S. Gorodkov, M. I. Gurevich, M. A. Kalugin, A. S. Kulakov, S. V. Marin, A. P. Novosel’tsev, D. S. Oleinik, A. V. Pryanichnikov, E. A. Sukhino-Khomenko, D. A. Shkarovskii, and M. S. Yudkevich, “The status of MCU-5,” Vopr. At. Nauki Tekh. Ser.: Fiz. Yad. Reaktorov, No. 4, 4–23 (2011).
“Best estimate safety analysis for nuclear power plants: Uncertainty evaluation,” Safety Reports Series, Vol. 52 (IAEA, Vienna, 2008).
H. Glaeser, “GRS method for uncertainty and sensitivity evaluation of code results and applications,” Sci. Technol. Nucl. Install. 2008, 798901 (2008). doi 10.1155/2008/798901
W. L. Oberkampf and C. J. Roy, Verification and Validation in Scientific Computing Cambridge Univ. Press, Cambridge, 2010).
D. P. Veprev, A. V. Boldyrev, S. Yu. Chernov, and N. A. Mosunova, “Development and validation of the BERKUT fuel rod module of the EUCLID/V1 integrated computer code,” Ann. Nucl. Energy 113, 237–245 (2018).
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Original Russian Text © V.M. Alipchenkov, A.V. Boldyrev, D.P. Veprev, Yu.A. Zeigarnik, P.V. Kolobaeva, E.V. Moiseenko, N.A. Mosunova, E.F. Seleznev, V.F. Strizhov, E.V. Usov, S.L. Osipov, V.S. Gorbunov, D.A. Afremov, A.A. Semchenkov, 2018, published in Teploenergetika.
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Alipchenkov, V.M., Boldyrev, A.V., Veprev, D.P. et al. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 2: Validation and Verification. Therm. Eng. 65, 627–640 (2018). https://doi.org/10.1134/S004060151809001X
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DOI: https://doi.org/10.1134/S004060151809001X