[en] In the frame of Tacis Project R2.01/99, which was running from 2003 to 2005, the bubble condenser system of Kola NPP (unit 3) was qualified at the integral test facility BC V-213. Three LB LOCA tests, two MSLB tests, and one SB LOCA test were performed. The appropriate test scenarios for BC V-213 test facility, modeling accidents in the Kola NPP unit 3, were determined with pretest calculations. Analysis of test results has shown that calculated initial conditions and test scenarios were properly reproduced in the tests. The detailed posttest analysis of the tests performed at BC V-213 test facility was aimed to validate the COCOSYS code for the calculation of thermohydraulic processes in the hermetic compartments and bubble condenser. After that the validated COCOSYS code was applied to NPP calculations for Kola NPP (unit 3). Results of Tacis R2.01/99 Project confirmed the bubble condenser functionality during large and small break LOCAs and MSLB accidents. Maximum loads were reached in the LB LOCA case. No condensation oscillations were observed.

[en] Procedure of construction-mounting as well as maintenance and start-up works at the third power unit of the Kola NPP is described. Main stages of WWER-440 power unit construction are given. Works on electric mounting and sealing are described in brief. Recommendations to engineering, constructing and maintenance and start-up personnel are given

[en] Analysis of TLOFW(Total Loss of Feed-Water) has been performed using NPA based on MARS code with acceptance criteria considering design features for Kori 3 and 4. It is assumed that operator recognizes the event when the followings are satisfied : the loss of auxiliary feedwater, steam generator dryout, and pressurizer PORV automatic opening. Acceptance criteria are as follows : First, all the three PORV flow paths and all the two SI pumps have sufficient capacity to prevent core uncovery with two feet margin during a TLOFW event when all PORV flow paths are manually opened at 30minutes after the operator recognizes TLOFW. Second, two of three PORV flow paths in conjunction with one-of-two SI pumps have sufficient capacity to prevent core uncovery with two feet margin following a TLOFW event when two PORV flow paths are manually opened at recognition point of TLOFW. The analysis results show that the active core is covered with more than 2ft margin for the two cases, respectively

[en] The organizational principles of the constructional and mounting works in the Kola Nuclear Power Station are presented. The role of the technical preparation is emphasized. Large working capacity has been centralized. Large scale prefabrication was introduced into the mounting work. The operational organization played an important role in the organization of preoperational work. (R.J.)

No abstract available

[en] A description is given of the universal procedure for calculation of fast neutron fluence (FNF) on WWER vessels. Approbation of the calculation procedure was carried out by comparing the calculation results for this procedure and measurements on the outer surface of the WWER-440 and WWER-1000 vessels. In addition, an estimation of the uncertainty of the settlement procedure was made in accordance with the requirements of regulatory documents. The developed procedure is applied at Kola NPP for independent fast neutron fluence estimates on the WWER-440 reactor vessels when planning core loads taking into account the introduction of new fuels. The results of the pilot operation of the procedure for calculating FNF at the Kola NPP were taken into account when improving the procedure and its application to the calculations of FNF on the WWER-1000 vessels

[en] Thermal-physical characteristics of fuel rods of two fuel assemblies which were operated within 5 - 8 and 5 - 9 core fuel loadings of the Unit 3 of the Kol'skaya NPP are calculated. They have achieved deep burnup during 4-year (> 46 Mwd/kg U) and 5-year (> 48 Mwd/kg U) fuel cycle. Fuel assemblies have been unloaded off the reactor and subjected to a post-irradiation testing. PIN-mod2 code originally designed for modelling of WWER fuel rod behaviour in a quasi-steady-state operation is used. The average fuel rod in the fuel assembly and the fuel rod with maximum burnup are selected. The preliminary comparison of the calculation results with those of the post-irradiation examination shows a satisfactory agreement. On the basis of the results obtained in the post-irradiation experiments an improvement of the model for calculation of fission gas release and creep of the cladding is planned. The results of the analysis performed indicate that the fuel rod completely preserves its working ability; fuel temperature does not exceed 1300^o C; fission gas release does not exceed 4%; maximum gas pressure inside the cladding at the end of campaign does not exceed 2 MPa. 2 tabs., 11 figs., 5 refs

No abstract available

[en] Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

[en] In this paper the following achieved results are discussed: 1) Innovation of PIN code for HiBu domain; 2) Innovation of FRAS code for burnt fuel transient simulation; 3) Development and optimization of mechanistic Fission Gas Release (FGR) model; 4) Design of PIN to FRAS (PIN2FRAS) interface; 5) Testing and validation runs of PIN2FRAS coupling on tentative RIA like accident superposed to the KOLA NPP (FA 222, pin 007) irradiation history; 6) Testing of PIN with built in diffusion FGR model; 7) Testing of PIN2FRAS with/without diffusion FGR model