Effect of statistical characteristics of fuel pin bundle on evaluation of temperature in the core of sodium cooled fast reactor
UDC: 621.039.526: 621.039.534.23
The article is devoted to optimization of analytical models used for the evaluation of fast reactor (FR) fuel element cladding temperature with the aim to increase the fidelity of the results obtained.
The rated value of the fuel element cladding temperature at the core outlet in FR fuel subassemblies is basically determined by the coolant temperature. However, calculation of the coolant temperature in the channels surrounding the fuel elements requires knowledge of the model of the fuel pin bundle arranged in a triangular grid. Distribution of the local flow rates and, hence, coolant temperature rise values in the channels over the pin bundle cross section is fully dependent on the arrangement of the fuel elements in the fuel subassembly wrapper taking into account the technological fitting gap.
Various hypothetical bundle models adopted for the calculation were studied and their drawbacks were identified. The difference in sodium temperature rise values for the different models is over 30 °С. The models are static, so they do not allow taking into account to the full extent radial and, particularly, axial heat and mass transfer that would decrease non-uniformity of the fuel element cladding temperatures.
A bundle model was developed based on the results of experimental studies of fuel element arrangement in fresh fuel subassemblies. It is demonstrated that formation of channel dimensions at the stage of manufacturing subassemblies with fuel pin bundles is a stochastic process following statistical laws. The relationships between characteristics of the fuel pin bundle required for calculation and the value of the technological fitting gap in the fuel subassembly were obtained by processing statistic data using the Weibull distribution.
There are two pin bundle models created taking into account statistics, one of which allows using the most probable statistically average dimensions of different type channels instead of hypothetical dimensions in the calculation of the rated value of sodium temperature rise at the design development stage. The second option makes it possible to evaluate the random dispersal of temperature rise values in the pin bundle taking into account axial and radial heat and mass transfer. This significantly increases the fidelity of evaluating the working temperatures of the hottest fuel elements.
- Usynin G.B., Karabasov A.S., Chirkov V.A. Optimizatsionnye modeli reaktorov na bystryh nejtronah [Optimisation models for fast neutron reactors]. Moscow, Atomizdat Publ., 1981. 232 p. (in Russian)
- Kuzevanov V.S., Baklushin R.P. Teplovoj i gidravlicheskij raschyot aktivnyh zon reaktorov: Uchebnoe posobie po kursu «Teplogidravlicheskij raschyot reaktorov» [Calculation of thermal and hydraulic characteristics of reactor cores: Textbook on thermal and hydraulic analysis of reactors]. Obninsk, OIAE Publ., 1988. 64 p. (in Russian)
- Skok Zh. Teplovye i gidravlicheskie issledovaniya aktivnoj zony 250 M. Feniks. Avant-proekt. [Thermal and hydraulic studies on 250 M core. Phoenix. Avante-project]. Per. s fr. Moscow, Vsesoyuznyj centr perevodov nauchno-tekhnicheskoj literatury i dokumentov. Per. No. SR-2942, Publ.1977. 93 p.
- Miki K. Deformation analysis of fuel pins within the wire-wrap assembly of an LMFBR. Nuc. Eng. and Des. 1979, v. 52, pp. 371–382.
- Weber G., Cornet G. Thermohydraulic characteristics of SNR – fuel elements. Karlsruhe, FRG, 5 – 7 February 1978, JWGFR-29, pp. 29–58.
- Usynin G.B., Kusmartsev E.B. Reaktory na bystryh nejtronah. Pod red. F.M. Mitenkova. [Fast neutron reactors]. Moscow, Energoatomizdat Publ.1985. 288 p. (in Russian)
- Klyomin A.I. Strigulin M.M. Nekotorye voprosy nadyozhnosti yadernyh reaktorov [Some issues of reliability of nuclear reactors]. Moscow, Atomizdat Publ. 1968. 352 p. (in Russian)
- Kurbatov I.M., Tikhomirov B.B. Raschyot sluchajnyh otklonenij temperatur v aktivnoj zone reaktora: Preprint FEI-1090. [Evaluation of random departures of reactor core temperatures. Preprint FEI-1090]. Obninsk, FEI, Publ.1980. (in Russian)
- Zhukov A.V., Kirillov P.L., Matyukhin N.M., Sorokin A.P., Tikhomirov B.B., Ushakov P.A., Yur’ev Yu.S. (SSSR), Mantlik F., Gejna Ya., Shmid J., Shul’c V., Krett V. (ChSSR). Teplogidravlicheskij raschyot TVS bystryh reaktorov s zhidkometallicheskim ohlazhdeniem [Calculation of thermal and hydraulic characteristics of fuel subassemblies of liquid metal cooled fast reactors]. Moscow, Energoatomizdat Publ. 1985. 160 p. (in Russian)
- Bogoslovskaya G.P., Zhukov A.V., Poplavskij V.M., Sorokin A.P., Tikhomirov B.B., Ushakov P.A. Metod raschyota temperaturnogo polya v kassete tvelov bystrogo reaktora pri sluchajnom raspredelenii parametrov po metodu Monte-Karlo; Preprint FEI-1340. [Method of calculation of temperature profile of the fuel subassembly of fast reactor with random parameters distribution using Monte Carlo method. Preprint FEI-1340]. Obninsk, FEI Publ. 1982. (in Russian)
- Marbach J. Comportement d’un faisceau d’aiquilles Phenix sour irradiation. –In Irradiation Behaviour of Metallic Materials for Fast Reactor Core Components. Et Edite par J. Poirier et. I.M. Dupony – CEA – DMECH – B.P. № 2 – 91190 GIF – Sur – YIETTE, France, pp. 297 – 301. (in French)
- Cognet G. Evolution en fonction de l’irradiation du profil de temperature en sortie du faisceau d’assemblage dans le reacteur Rapsodie – experience TETACOUPLE. Seminaire CEA – GKAE – Octobre 86 – CEN Cadarache, France. (in French)
- Leteinturier D., Cartier L. Theoretical and Experimental Investigations of the Thermohydraulics of Deformed Wire – Wrapped Bundles in Nominal Flow Conditions. In ref. , pp. 254 – 261.