Core designing of a new type of TVS-2M FAs: neutronics and thermal-hydraulics design basis limits
Saeed GHAEMI, Farshad FAGHIHI
Core designing of a new type of TVS-2M FAs: neutronics and thermal-hydraulics design basis limits
One of the most important aims of this study is to improve the core of the current VVER reactors to achieve more burn-up (or more cycle length) and more intrinsic safety. It is an independent study on the Russian new proposed FAs, called TVS-2M, which would be applied for the future advanced VVERs. Some important aspects of neutronics as well as thermal hydraulics investigations (and analysis) of the new type of Fas are conducted, and results are compared with the standards PWR CDBL. The TVS-2M FA contains gadolinium-oxide which is mixed with UO2 (for different Gd densities and U-235 enrichments which are given herein), but the core does not contain BARs. The new type TVS-2M Fas are modeled by the SARCS software package to find the PMAXS format for three states of CZP and HZP as well as HFP, and then the whole core is simulated by the PARCS code to investigate transient conditions. In addition, the WIMS-D5 code is suggested for steady core modeling including TVS-2M FAs and/or TVS FAs. Many neutronics aspects such as the first cycle length (first cycle burn up in terms of MWthd/kgU), the critical concentration of boric acid at the BOC as well as the cycle length, the axial, and radial power peaking factors, differential and integral worthy of the most reactive CPS-CRs, reactivity coefficients of the fuel, moderator, boric acid, and the under-moderation estimation of the core are conducted and benchmarked with the PWR CDBL. Specifically, the burn-up calculations indicate that the 45.6 d increase of the first cycle length (which corresponds to 1.18 MWthd/kgU increase of burn-up) is the best improving aim of the new FA type called TVS-2M. Moreover, thermal-hydraulics core design criteria such as MDNBR (based on W3 correlation) and the maximum of fuel and clad temperatures (radially and axially), are investigated, and discussed based on the CDBL.
TVS-2M FAs / core design basis limits / VVER-1000 / analysis / mixture of uranium-gadolinium oxides fuels / thermal-hydraulics / PARCS / WIMS-D5
[1] |
Bibilashvili Yu K, Kuleshov A V, Mikheev E N,
|
[2] |
Une K. Thermal expansion of UO2-Gd2O3 fuel pellets. Journal of Nuclear Science and Technology, 1986, 23(11): 1020–1022
CrossRef
Google scholar
|
[3] |
Lysikov A V, Kouleshova V, Novikova E A. Results of thermal-physical and mechanical property investigations of modified uranium-gadolinium oxide doped fuel. In: Russian Scientific Conference Materials for Nuclear Technics. Zvenigorod, Russia, 2007, 22–44
|
[4] |
Nivikov V, Dolgov A, Molchanov V. In: TopFUEL Conference. Wurzburg, Germany, 2003
|
[5] |
Kosourov E, Oleksuk D, Pavlovichev A, Shecherenko A. Optimization of gadolinium distribution in FAs for WWER-1000. In: 10th International Conference on WWER Performance, Modelling and Experimental Support. Sandanski, Bulgaria, 2013
|
[6] |
Oka Y, Madarame H, Uesaka M. An advanced course in nuclear engineering: nuclear reactor design. Japan: Springer Press, 2014
|
[7] |
Porhemmat M H, Hadad K, Faghihi F. PARCS cross-section library generator; part one: development and verification. Progress in Nuclear Energy, 2015, 78: 155–162
CrossRef
Google scholar
|
[8] |
Safaei Arshi S, Mirvakili S M, Faghihi F. Modified COBRA-EN code to investigate thermal-hydraulics analysis of the Iranian VVER-1000 core. Progress in Nuclear Energy, 2010, 52(6): 589–595
CrossRef
Google scholar
|
[9] |
Lunin G L, Novikov A N, Pavlov V I, Pavlovichev A M. Development of four-year fuel cycle based on the advanced fuel assembly with uranium-gadolinium fuel and it simple mentation to the operating VVER-1000 units. In: Proceeding of the 10th Symposium of AER. Moscow, Russia, 2000, 1:75–95
|
[10] |
Amin Mozafari M, Faghihi F. Design of annular fuels for a typical VVER-1000 core: neutronic investigation, pitch optimization and MDNBR calculation. Annals of Nuclear Energy, 2013, 60: 226–234
CrossRef
Google scholar
|
[11] |
Mirvakili S M, Faghihi F, Khalafi H. Developing a computational tool for predicting physical parameters of a typical VVER-1000 core based on artificial neural network. Annals of Nuclear Energy, 2012, 50: 82–93
CrossRef
Google scholar
|
[12] |
Bagheri S, Faghihi F, Nematollahi M R, Behzadinejad B. Assessment of thermal hydraulics parameters of the VVER-1000 during transient conditions. International Journal of Hydrogen Energy, 2016, 41(17): 7103–7111
CrossRef
Google scholar
|
[13] |
Lamarsh J R. Introduction to Nuclear Reactor Theory. New York University: Addison Wesley Publishing Company, 1966
|
[14] |
Tong L S, Weisman J. Thermal Analysis of Pressurized Water Reactors. United States Department of Energy, Office of Scientific and Technical Information, OSTI Identifier: 5856843, 1979
|
[15] |
Novikov V V, Mikheev E N, Lysikov A V,
|
[16] |
Balestrieri D. A study of the UO2/Gd2O3 composite fuel, IAEA Series. https://inis.iaea.org/search/search.aspx?orig_q=RN:290547-47
|
[17] |
Todreas N E, Kazimi M S. Nuclear Systems I: Thermal Hydraulics Fundamentals. CRC Press, 2001
|
[18] |
Faghihi F, Fadaie A H, Sayareh R. Reactivity coefficients simulation of the Iranian VVER-1000 nuclear reactor using WIMS and CITATION codes. Progress in Nuclear Energy, 2007, 49(1): 68–78
CrossRef
Google scholar
|
[19] |
Downar T, Xu Y, Seker V,
|
[20] |
Downar T, Xu Y, Seker V,
|
[21] |
Lunin G, Novikov A, Pavlov V,
|
[22] |
Atomic Energy Organization of Iran. Album of Neutron and Physical Characteristics of the 1st Loading of Boushehr Nuclear Power Plant. BNPP FSAR, 2003
|
[23] |
Faghihi F, Saidi nezhad M. Two safety coefficients for a typical 13 × 13 annular fuel assembly. Progress in Nuclear Energy, 2011, 53(3): 250–254
CrossRef
Google scholar
|
[24] |
Erfani Nia A, Faghihi F, Hadad K. Prompt and power reactivity coefficients for the next generation VVER-1000 reactor including hexagonal assemblies and annular fuels. Progress in Nuclear Energy, 2012, 61: 41–47
CrossRef
Google scholar
|
[25] |
Driscoll M J, Downar T J, Pilat E E. The Linear Reactivity Model for Nuclear Fuel Management. La Grange Park, Illinois: American Nuclear Society, 1990
|
/
〈 | 〉 |