Izvestiya vuzov. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

Results of Validation and Cross-Verification of the ROK/B Design Code on the Problem of Loss of Cooling in the Spent Fuel Pool

12/08/2021 2021 - #04 Global safety, reliability and diagnostics of nuclear power installationss

Sledkov R.M. Karnaukhov V.E. Stepanov O.E. Bedretdinov M.M. Chusov I.A.

DOI: https://doi.org/10.26583/npe.2021.4.08

UDC: 621.039.58

The procedures of validation and cross-verification of the newly developed computational code ROK/B are described. The main problem solved using the ROK/B code is the substantiation by calculation of the coolant density in the spent fuel pool (SFP) (untight reactor) and the temperature regime of the fuel assemblies during a protracted shutdown of the cooling systems (break in the supply of cooling water). In addition to the above, it is possible using the computational code ROK/B to carry out calculation of an accident with the discharge of the coolant from the SFP with simultaneous long-duration shutdown of the cooling systems.

The ROK/B computational code allows carrying out calculations for various types of designs of the fuel assemblies and VVER reactors, in particular, VVER-1000, VVER-1200 and VVER-440 power units with single- and two-tiered fuel assemblies arrangement, with clad pipes in racks (for compacted assemblies storage) and pipes without cladding, with cased assemblies and caseless ones.

During fuel reloading, a high level of the coolant is maintained, which makes it possible to do “wet” transportation of the assemblies from the reactor to the SFP. The mathematical model for heat and mass transfer calculation, including the boiling coolant model, implemented in the ROK/B code, includes: the motion equation, equations for calculating the enthalpy along the height of the fuel section of a fuel assembly with natural circulation of coolant within the channel containing the fuel assembly (lifting section) and in the inter-channel space (lowering section), the equation of mass balance between the channels of the racks with assemblies and in the inter-assembly space and the amount of evaporated (and outflowed) water, the heat balance equation for a fuel rod in a steam environment. The system of equations is supplemented by closing relations for calculating the thermal physics properties of water and steam, fuel and cladding materials, as well as the coefficients of heat transfer from the wall to the steam, hydraulic resistance and density of the steam-water mixture in the channels, and the heat released in the reaction of steam with zirconium.

Validation of the computational code was carried out on the basis of the data of the ALADIN experiment performed by German specialists and the data of OKB Gidropress JSC. Cross-verification of the ROK/B computational code was carried out in comparison with the results of calculation using the KORSAR/GP and SOCRAT/B1 codes. Based on the results of the validation, it was concluded that the deviation of the ROK/B results from the experimental data is not more than 2 – 10% (10% for the option with a fuel rod power of 20 W). Based on the results of cross-verification, it was concluded that the discrepancy between the ROC/B results and the calculation results for the KORSAR/GP and SOCRAT codes is not more than 0.5% (for SOCRAT/V1) and less than 10% (for KORSAR/GP).


  1. Kritsky V.G., Berezina I.G., Kalinkin V.I., Tikhonov N.S., Kozlov Yu.V., Razmashkin N.V., Shafrova N.P. «Atomic Renaissance» and Prospects of SNF Treatment. Bezopasnost’ Okruzhayuschey Sredy. 2008, no. 1, pp. 68-71 (in Russian).
  2. Gagarinskiy A.Yu. RW Management in the Nuclear Energy Strategy of Russia. Energiya: Ekonomika, Tekhnika, Ekologiya. 2014, no. 7, pp. 2-9 (in Russian).
  3. Safutin V.D., Simanovsky V.M. Tikhonov N.S. Transportation and Storage of Spent Fuel assemblies. In Encyclopedia «Machine Building», Vol. IV «Machine2Building of Nuclear Engineering», Book 2. Moscow. Mashinostroenie Publ., 2005, pp. 280-300 (in Russian).
  4. Phenomena Identification and Ranking Table. Priorities for Loss-of-Cooling and Loss-of-Coolant Accidents in Spent Nuclear Fuel Pools. Nuclear Safety and Regulation. OECD 2018 NEA No. 7443. Available at: https://www.oecd-nea.org/upload/docs/application/pdf/2019-12/7443-pheno_id_rank_table.pdf (accessed Jun. 10, 2012).
  5. Wang D. et al. Study of Fukushima Daichi Nuclear Power Station Unit 4 Spent Fuel Pool. Nuclear Technology. 2012, v. 180, pp. 205 - 215; DOI: https://doi.org/10.13182/NT12-A14634 .
  6. Hydrodynamic Characteristics of Atmospheric Pressure Air Mixture in a Vertical Beam at High Weighted Flow Rates and Elevated Temperature. Report 76-0-011, OKB «Gidropress». Podolsk. OKB «Gidropress» Publ., 1962, – 72 p. (in Russian).
  7. Partmann Christine, Schuster Christoph, Hurtado Antonio. Experimental investigation of the thermal hydraulics of a spent fuel pool under loss of active heat removal conditions. Nuclear Engineering and Design. 2018, v. 330, pp.480-487; DOI: https://doi.org/10.1016/j.nucengdes.2018.02.023 .
  8. Petkevich I.I., Uvakin M.A. Application of the LINQUAD Program for Analysis of Uncertainties in Calculations of Regime with a Rupture of Steam Line at the AES-2006 Unit According to the KORSAR/GP code. VANT. Ser: Fizika Yadernykh Reaktorov. 2013, iss. 2, pp. 51-60 (in Russian).
  9. Pantyushin S.I., Sorokin Yu.S. Comparison of the Codes SOCRAT/V1, KORSAR/GP and TECH-M-97 in the Analysis of the Initial Stage of a Severe Accident at the VVER-1200 RP. Proc. of the VI2th International Scientific and Technical Conference «Ensuring the Safety of NPP with VVER». Podolsk, May 26229, 2009. Podolsk. OKB «Gidropress» Publ., 2009, pp. 71-79 (in Russian).
  10. Alexandrova O.L., Barabanov R.A., Dyanov D.Yu., Kosarim S.S., Naumov A.O., Spiridonov V.F., Filimonkin E.A., Tsiberev K.V. LOGOS software package. A Finite Element Method for Calculating Problems of Static Strength of Structures Taking Account of Effects of Physical and Geometrical Nonlinearity. VANT. Ser: Matematicheskoe Modelbrovanie Fizicheskikh Processov. 2014, iss. 3, pp. 3-17 (in Russian).
  11. Klemin A.I., Polyanin L.N., Strigulin M.M. Thermal2Hydraulic Calculation and Thermal Reliability of Nuclear Reactors. Moscow. Atomizdat Publ., 1980, 327 p. (in Russian).
  12. Sledkov R.M., Stepanov O.E. Cross-Verification of the ROK2 Program for the Problem with Loss of Cooling of the VVER-1000 Main Circuit. VANT. Ser: Fizika Yadernykh Reaktorov. 2017, iss. 1, pp. 32-36 (in Russian).
  13. Sledkov R.M., Stepanov O.E., Galkin I.Yu., Strebnev N.A. Correlation analysis of the true volumetric steam content for problems with loss of cooling of the spent fuel pool. Teploenergetika. 2017, no. 1, pp. 20-24; DOI: https://doi.org/10.1134/S004036361610009X . (in Russian).
  14. Improvement of Serial Reactor Unit with VVER21000 to Ensure Operational Safety and Reliability. Study of the Temperature Regime of the Cassette from the Simulated Fuel Elements at the Stage of Re2Fueling. Report on Research Work (Final 341-O-044), GKAE OKB «Gidropress». Podolsk. GKAE OKB «Gidropress» Publ., 1985, 112 p. (in Russian).
  15. Research Program. DENOPI Project. Spent Fuel Pool Loss-of-Cooling and Loss-of-Cooling Accident. 2021. Available at: https://www.irsn.fr (accessed Jun. 10, 2021).

spent fuel pond calculation code ROK/B KORSAR/GP SOCRAT/V1 loss of cooling VVER swelling validation fuel rod fuel assembly